kata pengantariconets.org/wp-content/uploads/2017/10/upload_buku...(issmm) & robotic contest...

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KATA PENGANTAR

Seminar Nasional Teknologi Energi Nuklir (SENTEN) yang ke-4 dan 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS-2017) bersama-sama dengan 4th International Symposium on Smart Material and Mechatronics (ISSMM) & Robotic Contest dengan tema ”Kontribusi Teknologi Energi Nuklir Bagi Kemandirian dan Keberlanjutan Pembangunan Nasional”, diselenggarakan oleh Pusat Teknologi dan Keselamatan Reaktor Nuklir (PTKRN - BATAN), Pusat Kajian Sistem Energi Nuklir (PKSEN-BATAN), Pusat Teknologi Bahan Bakar Nuklir (PTBBN-BATAN), Pusat Teknologi Limbah radioaktif (PTLR-BATAN), Pusat Teknologi Bahan Galian Nuklir (PTBGN-BATAN) dan bekerjasama dengan Fakultas Teknik - Universitas Hassanudin Makassar, Himpunan Masyarakart Nuklir Indonesia (HIMNI), Himpunan Peneliti Indonesia cabang BATAN dan Pemda Provinsi Sulawesi Selatan. Seminar ini diharapkan dapat menjadi ajang tukar menukar informasi antara peneliti, akademisi dan pemerhati terkait dengan penelitian dan pengembangan iptek energi nuklir dan aspek pendukungnya di Indonesia.

Dalam kegiatan ilmiah ini akan ditampilkan sesi Pleno dengan menampilkan keynote speaker yang berasal dari BATAN, Universitas Hasanuddin, BAPETEN, Okayama University Jepang, dari Division of Nuclear Power IAEA (Dr. Frederik Reitsma), Rosatom (Rusia) dan Tim Basic Engineering Desain – RDE BATAN.

Setelah melalui seleksi dan evaluasi oleh Dewan Editor, panitia memutuskan Panitia SENTEN-2017 menerima 68 makalah teknis yang berasal dari BATAN yaitu PTKRN, PKSEN,PTBGN, PTBBN, PTLR, PSMN, PTKMR, PSTA, PSTNT dan STTN, BAPETEN, UNHAS), sedangkan untuk ICoNETS-2017/ISSMM-2017 menerima sebanyak 67 makalah teknis dari berbagai instansi, dari beberapa negara yaitu Jepang, Korea, dan Australia, yang dipresentasikan dalam bentuk presentasi oral maupun poster serta tambahan 2 makalah undangan dari KINGS Korea dan Universitas Okayama Jepang.

Pada kesempatan ini, Panitia mengucapkan terimakasih yang sebesar-besarnya kepada berbagai pihak yang telah membantu terselenggaranya Seminar Nasional dan Internasional ini.

Jakarta, Oktober 2017

Panitia

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PREFACE

The 4th National Seminar on Nuclear Energy Technology (Seminar Nasional Teknologi Energi Nuklir, SENTEN) 2017 and the 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS-2017) in conjunction with the 4th International Symposium on Smart Material and Mechatronics (ISSMM) & Robotic Contest with the theme “Nuclear energy for a realistic approach of the national energy security and long term sustainability”, by Center for Nuclear Reactor Technology and Safety (PTKRN – BATAN), Center for Nuclear Energy System Assesment (PKSEN – BATAN), Center for Nuclear Fuel Technology (PTBBN-BATAN), Center for Radioactive Waste technology (PTLR-BATAN), Center for Nuclear Mines Technology (PTBGN-BATAN) in cooperation with Engineering Faculty of Hasanuddin University, Makassar, Association of Indonesian Nuclear Society (HIMNI), Indonesia Rresearcher Union BATAN Branch (HIMPENINDO-BATAN) and South Sulawesi Province Local Government. The seminars are expected to give an opportunity for researchers, academicians, and observers in the field of nuclear energy science and technology to share and discuss their current researches and achievements.

In these events, plenary session will be conducted with important keynote speakers including Chairman of BATAN, Rector of Hasanuddin University and Chairman of BAPETEN, Nuclear Power Division of IAEA, Okayama University Japan, Rosatom Company, PT. Timah and BATAN - Basic Engineering Design team of Experimental Power Reactor (RDE).

After going to the selection and evaluation procedures by the Editorial Committee, it was decided that 68 and 67 technical papers coming from various institutions plus 2 invited papers are accepted to be presented in SENTEN-2017 and ICoNETS-2017/ISSMM-2017, respectively. Technical paper for SENTEN-2017 is coming from BATAN i.e.: PTKRN, PKSEN, PTBGN, PTBBN, PTLR, PSMN, PTKMR, PSTA, PSTNT, STTN, BAPETEN and UNHAS. The ICoNETS-2017/ISSMM-2017 received 67 technical papers from various institutions, from several countries such as Indonesia, Japan, Korea and Australia, presented in oral and poster presentations, as well as two invited papers from KINGS Korea and Japan Okayama University.

The presentations will be divided into oral and poster presentation. Most of the technical papers are contributed from BATAN and Hasanuddin University.

SENTEN-2017 and ICoNETS-2017 Committee express great gratitude to all participants, exhibitors, and sponsors for all the contributions by which we can have a successfully. Looking forward for your contributions in next SENTEN and ICoNETS/ISSMM.

Jakarta, October 2017 Committee

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DAFTAR ISI

TABLE OF CONTENTS

Hal./Pages

Kata Pengantar 2

Preface 3

Daftar Isi / Table of Contents 4

Seminar Nasional ke-4 Teknologi Energi Nuklir (SENTEN-2017) 5

The 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS)

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Ketentuan Persidangan Seminar Nasional ke-4 Teknologi Energi Nuklir, the 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS), The 4th International Symposium on Smart Material and Mechatronics (ISSMM)

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Rule of The Seminar 12

Jadwal Acara / Program Agenda 13

SENTEN/ICoNETS/ISSMM Parallel Sessions 15

Daftar Kumpulan Judul makalah SENTEN-2017 19

Paper Code & Title Listing of ICoNETS/ISSMM 2017 24

Kumpulan Abstrak SENTEN-2017 29

Collected Paper Abstracts of ICoNETS/ISSMM - 2017 64

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Seminar Nasional ke-4 Teknologi Energi Nuklir (SENTEN-2017)

PENDAHULUAN Indonesia sebagai negara berkembang yang mengarah menjadi negara maju diindikasikan dengan dominasi sektor industri dalam menunjang perekonomiannya. Peranan sektor industri dalam penggunaan energi selalu mendominasi dan terus meningkat. Oleh karenanya jaminan ketersediaan energi sangat menentukan keberlanjutan pembangunan industri. Sesuai Kebijakan Energi Nasional, PP No. 79 Tahun 2014 bahwa target kapasitas listrik terpasang pada tahun 2025 adalah 115 GWe, yang pada saat sekarang baru mencapai lebih kurang 50 GWe. Energi nuklir menjadi keniscayaan opsi untuk mengejar pemenuhan kebutuhan energi listrik dan mengingat kebutuhan kapasitas daya terpasang yang terus meningkat. Dari sisi kesiapan teknologinya, pembangunan PLTN yang merupakan implementasi pemanfaatan energi nuklir, selain membantu mengamankan pasokan listrik nasional juga memberikan leverage ekonomi dan industri nasional. Penyediaan listrik yang masif memberikan leverage ekonomi yang dapat langsung dirasakan. Selain itu, karaktersitik PLTN yang memiliki teknologi dan berstandar keselamatan yang tinggi memberikan dampak pada standar kualitas industri nasional yang tinggi pula dan peningkatan kapasitas sumber daya manusia (SDM), khususnya pada perguruan tinggi. Seminar ilmiah nasional ini merupakan salah satu sarana untuk membangun penguatan teknologi dan aspek SDM serta pengembagan industrinya. Seminar ini memfasilitasi para peneliti, praktisi, akademisi dan pemerhati serta pemangku kepentingan untuk bertukar informasi terkait pengembangan teknologi energi nuklir dalam menjawab tantangan pengembangan industri nasional dan peningkatan kapasitas Sumber Daya Manusia (SDM). Seminar Nasional Teknologi Energi Nuklir (SENTEN) ini merupakan pertemuan ilmiah tahunan yang pada tahun 2017 diselenggarakan di Makassar atas kerjasama Kedeputian Teknologi Energi Nuklir BATAN dengan Fakultas Teknik Universitas Hasanuddin, Makassar.

TUJUAN Menginformasikan berbagai hasil kajian/litbang teknologi nuklir dan iptek pendukungnya. Memfasilitasi para peneliti, praktisi, akademisi dan pemerhati serta pemangku kepentingan

untuk bertukar informasi terkait pengembangan teknologi energi nuklir dalam menjawab tantangan pengembangan industri nasional dan peningkatan kapasitas Sumber Daya Manusia (SDM).

RUANG LINGKUP Teknologi Reaktor dan Keselamatan Nuklir Teknologi Pengelolaan Limbah Radioaktif Teknologi Bahan Galian Nuklir Teknologi Bahan Bakar Nuklir Pengkajian Sistem Energi Nuklir Bidang Keteknikan: Teknik Mesin, Teknik Perkapalan, Teknik Arsitektur, Teknik Elektro,

Teknik Sipil dan Teknik Geologi. Bidang Sosbud, Keamanan dan Ekonomi terkait Ketenagaan Nuklir

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PELAKSANAAN Hari/tanggal : Kamis, 12 Oktober 2017 Waktu : 08.00 sd. selesai Tempat : Fakultas Teknik, Universitas Hassanudin Kampus Gowa, Makassar, Sulawesi Selatan

DEWAN EDITOR SENTEN-2017 Dewan Editor 1. Dr. Ir. P. Made Udiyani, M.Si. (BATAN) 2. Dra. Heny Susiati, M.Si. (BATAN) 3. Dr. Mulya Juarsa, S.Si., MESc. (BATAN) 4. Dr. Ir. Hendro Tjahjono, DEA (BATAN) 5. Dr. Sigit Santoso, M.Eng. (BATAN) 6. Ir. D. T. Sony Tjahyani, M.Eng. (BATAN) 7. Dr. Arya Adhyaksa Waskita, S.Si., M.Si. (BATAN) 8. Drs. Sahala M. Lumban Raja (BATAN) 9. Yuliastuti, S.Si., M.Si. (BATAN) 10. Dr. Dede Sutarya, ST., MT. (BATAN) 11. Ngadenin, ST. (BATAN) 12. Drs. M.Najib (BATAN) 13. Dr. Ir. Budi Setiawan, M.Eng. (BATAN) 14. Siti Hijraini Nur, ST., MT (UNHAS) 15. Dr. Faizal Arya Samman (UNHAS) 16. Dr. Jalaluddin (UNHAS) 17. Dr. Arya Subahia (UNUD) 18. Dr.-Ing. Sihana (UGM)

Website: http://www.batan.go.id/index.php/id/ ptkrn-id/senten/senten-2017

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THE 2ND INTERNATIONAL CONFERENCE ON NUCLEAR TECHNOLOGIES AND SCIENCES (ICoNETS-2017)

INTRODUCTION Following the previous successful 1st International Conference on Nuclear Technologies and Sciences of 2015, ICONETS 2015, five research centers under the Deputy of Nuclear Energy Technology – National Nuclear Energy Agency in collaboration with Indonesia Researcher Community (Himpunan Peneliti Indonesia- Himpenindo), also supported by Universitas Hasanuddin (Hasanuddin University) organizes the second International Conference on Nuclear Technology and Sciences in conjunction with International Symposium on Smart Material Mechathronics (ISSMM) with theme :

“Nuclear energy for a realistic approach of the national energy security and long term sustainability”.

This conference is one of the forums which are expected to strengthen the nuclear program as well as to intensively discuss the future solution, strategy and development. This conference which is held regularly every two years, can be a melting pot of the stake holders which are concern with nuclear energy developments. The conference covers all the aspects of the fuel cycle in nuclear power plants, from policy to nuclear waste managements. This year conference is expected to involve greater regional (Asia Pacific) and international nuclear community contributions. This year, topic conference is expected to address the energy sustainability and security in the long term future by means of nuclear energy contribution. This contribution is taken for better approaches in conference theme in providing solution internationally while giving the best enforcement of Indonesia strategy program solution. In conjunction with the International Seminar, the annual national seminar, the 4th National Nuclear Energy Technology Seminar (SENTEN-2017) are also held to enhance wide national participations as well as to disseminate the current state efforts of the Indonesia Nuclear Program.

Regarding Indonesian cases, the current government strategy to elaborate all possible energy resources in answering the future demand energy projection not only requires fossil fuel power but also the new and renewable resources. Relaying on the fossil fuel as the major contributors, as the Indonesian example to more than 60% generation, may jeopardize future energy security and unable to guarantee the strong sustainability of energy generation frame. Following this, president of Indonesia stated that nuclear must be considered as the one of solution and cannot be floated. Addressing this direction, government of Indonesia prepares the construction of experimental power reactor to multiply the public acceptance as well as to improve the capacity building in terms of technology, industry, and human resources readiness. This strategy, learning from the other previous embarking countries, can be a certain step for the better implementation of nuclear energy in the future for the embarking countries.

PURPOSE AND OBJECTIVES The conference is being covered to discuss the national and international experiences in strengthening nuclear programs for the sustainable and security of energy. This forum will be an ideal for : 1. Nuclear technology and science exercises 2. Review the current status of the nuclear energy technology and sciences

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3. Review the current status and efforts of the policy, program, strategy and regulation of the nuclear power plant embarking countries as well as learn the best practice of previous well established countries

4. Reaffirm and support the government, industry and university in nuclear technology program including other related activity such as public acceptance, infrastructure as well as the nuclear applications in various field

5. Raise the awareness of nuclear technology contributions for public, local, national and regional governments

6. Network strengthening with technology provider/vendor.

This conference expects participants from government policy makers, regulatory body, technical support experts, technology vendor and academician from universities. The conference will be three parts: ministerial and policy maker forums, technical forums, and technology vendor forums. TOPICS The conference welcomes the contribution papers on all aspects of nuclear energy technology, nuclear application sciences, policy and regulation, both academic and practice-based papers. The conference will consist of plenary session which covers ministerial and policy makers, industrial session, and topical sessions both oral and poster contributions. The topics of interests include but are not limited to, the following: 1. Nuclear Reactor Technology and Safety (water-cooled, high temperature, and advance

reactors) Neutronics for Research Reactor and Power Reactor Reactor Physics Technology Thermal Hydraulics for Reactor Safety Aging Management for Reactor Material and Reactor Integrity Nuclear Safety Culture

2. Radioactive Waste Technology Radioactive Waste Management Waste Processing and Storage Technology Development of Waste Facility Technology

3. Nuclear Mines Technology Exploration Field on Nuclear Geology Inventory of Nuclear Ore and Hydrogeology Technology of Mining and Processing Nuclear Ore Environmental and Work Control Safety

4. Nuclear Fuel Technology Nuclear Fuel Fabrication Technolog Radio-metallurgy Testing Technique Development Nuclear Fuel Facility Management Nuclear Material Accountancy and Work Safety Monitoring

5. Assessment of Nuclear Energy System Nuclear Energy Infrastructure Assessment NPP Site and Environmental Assessment

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National Energy Planning Assessment by Nuclear Option 6. Other Related Nuclear Energy Technology and Science Activities including Nuclear

Applications Technology, Security, Social, Economics, National Policy, Regulatory and Licensing, Government Planning and Human Resources Development.

VENUE Engineering Faculty Building, Hasanuddin University, Gowa Campus, Makassar Makassar, October 12, 2017. EDITORIAL BOARDS OF ICoNETS-2017 Editorial Boards: Dr. Geni Rina Sunaryo, M.Sc. Dr. Julwan Hendry Purba, ST., M. App.IT. Reviewers 1. Prof. Dr. Djarot S. Wisnubroto (BATAN) 2. Prof. Dr. Nesimi Ertugrul (UoA, Australia) 3. Prof. Dr. M. Wihardi Tjaronge (UNHAS Makassar) 4. Prof. Dr. Akio Gofuku (Okayama Univ., Japan) 5. Prof. Dr. Nguyen Trung Tinh (TIC-VARNS, Vietnam) 6. Prof. Dr.-Ing. Nandy Putra (UI, Depok) 7. Prof. Dr. Jung Jae-cheon (KINGS, Korea) 8. Prof. Dr. Ridwan (BATAN) 9. Prof. Dr. Ir. Dedi Priadi, DEA. (UI, Depok) 10. Dr. Hadid Subkhi (IAEA) 11. Mr. Frederik Reitsma (IAEA)

12. Dr. Jim Kuijper (NRG, Netherlands) 13. Dr. Mark Mitchell (PBMR, South Africa) 14. Dr. Mike Davies (AFW, UK) 15. Dr. Kunihiko Nabeshimaa (JAEA-Japan) 16. Dr. Sun Jun (Tsinghua University, China) 17. Dr. Phongpaeth Pengvanich (CU, Thailand) 18. Dr. Sidik Permana (ITB Bandung) 19. Dr. Alexander Agung (UGM, Yogyakarta) 20. Dr. Deendarlianto (UGM, Yogyakarta) 23. Rafiuddin Syam, Ph.D. (UNHAS, Makassar) 24. Dr. Wayan Nata Septiadi (UNUD, Bali)

IMPORTANT DATES Full paper submission : April 25, 2017 Notification of review report : July 25, 2017 Final paper submission : August 26, 2017 Final Registration : September 16, 2017 Event : October 12, 2017

Website http://iconets.org/

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KETENTUAN PERSIDANGAN Seminar Nasional Teknologi Energi Nuklir (SENTEN-2017),

The 2th International Conference on Nuclear Technologies and Sciences (ICoNETS-2017) The 4th International Symposium on Smart Material and Mechatronics (ISSMM)

A. Sidang Pleno, di Gedung Fakultas Teknik, Universitas Hasanuddin, Makassar 1. Sidang pleno akan menyajikan 7 makalah utama, yang akan disampaikan oleh

Pemakalah Kunci. 2. Seluruh peserta diharapkan hadir dan berperan aktif dalam Sidang Pleno. B. Sidang Presentasi Lisan/Oral, di Gd. Fakultas Teknik Universitas Hasanuddin. 1. Sidang Presentasi Lisan/Oral akan dilakukan secara paralel pada 4 ruangan yaitu

Room#1, Room#2, Room#3 dan Room#4. 2. Masing-masing Ruang/Group akan dipimpin oleh seorang Moderator sebagai Ketua

Sidang. Para Penyaji makalah diharapkan sudah menyerahkan atau mengkopi file presentasi ke komputer yang telah disediakan menjelang sesi persidangan.

3. Pada Sesi-1, alokasi waktu diberikan kepada 4 Penyaji makalah untuk melakukan presentasi masing-masing selama 15 menit termasuk diskusi/tanya jawab, kecuali 2 invited speaker diberi waktu presentasi 30 menit termasuk diskusi/tanya jawab.

4. Pada Sesi-2 alokasi waktu diberikan kepada 4 Penyaji makalah untuk melakukan presentasi masing-masing selama 15 menit termasuk diskusi/tanya jawab. Moderator akan mengontrol penggunaan waktu oleh tiap Penyaji/Presenter.

5. Dalam sesi tanya jawab, penanya dimohon menuliskan pertanyaan pada lembar pertanyaan, dan menyerahkan kepada Moderator. Selanjutnya Panitia akan menyerahkan lembar pertanyaan tersebut kepada Pemakalah, dan Pemakalah wajib menuliskan jawaban, dan menyerahkan kembali kepada Panitia. Sesi diskusi dan tanya jawab akan dimuat sebagai pelengkap makalah, pada Prosiding.

C. Sidang Presentasi Poster 1. Pembuatan Poster (SENTEN-2017 dan ICoNETS-2017) oleh peserta Seminar (tidak

dikoordinir panitia). 2. Isi poster dibuat dan dicetak oleh penulis sendiri sesuai format panitia dalam bentuk

X-Banner. 3. Pada saat sesi presentasi poster berlangsung, pemakalah diharapkan stand-by

menungu poster makalahnya masing-masing untuk menjawab pertanyaan yang disampaikan oleh peserta seminar.

4. Peserta Seminar diharapkan aktif bertanya pada sesi Sidang Presentasi Poster. Jika ada pertanyaan terkait dengan suatu makalah yang diposterkan, pemakalah wajib menjawab secara lisan, dan mencatat pertanyaan dan jawaban pada lembar

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tanya jawab yang telah disediakan. Selanjutnya pemakalah wajib menyerahkan lembar tanya jawab tersebut ke Panitia.

5. Sebagaimana Sidang Presentasi Lisan, lembar tanya jawab pada Sidang Presentasi Poster akan digunakan sebagai pelengkap makalah pada Prosiding.

D. Prosiding 1. Makalah yang dipresentasikan pada Seminar dan memenuhi syarat penerbitan,

akan dimuat pada Prosiding Seminar Nasional Teknologi Energi Nuklir (SENTEN) dengan Nomor ISSN 2355-7524.

2. Sesuai ketentuan Peneliti LIPI tentang penerbitan prosiding terkait dengan penilaian angka kredit Jabatan Fungsional Peneliti, buku prosiding akan diterbitkan setelah seminar, dan ditargetkan sekitar 2 (dua) bulan setelah pelaksanaan seminar.

3. Makalah yang dipresentasikan di ICoNETS/ISSMM akan dipublikasikan dalam proses terpisah pada perusahaan penerbitan IOP (Institute of Physics).

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RULE OF THE SEMINAR Seminar Nasional Teknologi Energi Nuklir (SENTEN-2017)

International Conference on Nuclear Technologies and Sciences (ICoNETS-2017) International Symposium on Smart Material and Mechatronics (ISSMM-2017)

A. Plenary Session, at Engineering Faculty of Hasanuddin University Building. 1. Plenary session includes one Panel Discussion and 7 presentations by keynote

speakers. 2. All of the audience should attend and actively involve in the Plenary Session. B. Oral Session, at Engineering Faculty of Hasanuddin University Building. 1. Oral Session of SENTEN/ICoNETS/ISSMM will be held in parallel in 4 different rooms. 2. Each rooms will be chaired by a Moderator. The speaker should give or copy the

presentation file into the prepared computer in each room prior to their session. 2. In a session, in the beginning each the speaker in that session give a 15 minutes

including presentation and discussion time. 3. Moderator should manage the presentation time of the speakers. 4. Complete documentations of question and answer in each session will be provided in

the proceedings. Questions from audiences and responses from the speakers should be collected to the committee. Committee will distribute question and answer forms in each session.

C. Poster Presentation Session 1. The author should make the poster based on specified format given by the committee

in the form of X-Banner type, as shown in Website. Printing of poster for SENTEN/ ICoNETS/ISSMM will not be organized by Committee.

2. At the poster presentation session, each author should stand by in the poster site for answering questions and discussions related to their poster.

3. All seminar attendants should actively involve by giving questions or discussions at the Poster Presentation Session. Author should write the questions and their answers in Q&A Form then collect it to the committee. This Q&A documentation will be included in the proceeding of the seminar.

D. Proceeding 1. Presented papers in the SENTEN-2017 which fulfill the publication requirements will

be published in the Prosiding Seminar Nasional Teknologi Energi Nuklir (SENTEN) with ISSN No. 2355-7524.

2. Following LIPI's rule on proceeding publication, it will be published after the seminar. The committee schedules the proceeding to be published around 2 months after the seminar.

3. Presented papers in the ICoNETS/ISSMM will be published on a separate proceeding which is IOP (Institute of Physics) publishing company.

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JADWAL ACARA / PROGRAM AGENDA Seminar Nasional ke-4 Teknologi Energi Nuklir (SENTEN-2017) &

The 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS-2017) The 4th International Symposium on Smart Material and Mechatronics (ISSMM-2017)

12 October 2017 Makassar, Sulawesi Selatan – Indonesia

Thur

sday

, 12 O

ctob

er 20

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08.00-08.30 Registration

08.30-08.40 Welcome Speech

by Prof. Dr. Dwia Aries Tina Pulubuhu, M.A Rector of Hasanuddin University

08.40-08.50 Welcome Speech and Opening Address

by Prof. Dr. Djarot Sulistio Wisnubroto Chairman of National Nuclear Energy Agency (BATAN)

08.50-10.20

Plenary I Moderator: Dr. Jalaluddin (UNHAS)

Sekretaris: Dr. Syaiful Bakhri (BATAN) Keynote Speech I

The Nuclear Technology Innovation in Supporting the Future National Energy Security

by Prof. Dr. Djarot Sulistio Wisnubroto Chairman of National Nuclear Energy Agency (BATAN)

Keynote Speech II The Regulatory Body Readiness for the Implementation of the first NPP in

Indonesia by Prof. Dr. Jazi Eko Istiyanto

Chairman of Nuclear Energy Regulatory Agency (BAPETEN) Keynote Speech III

The University Contribution for Nulcear Technology Application in East Region of Indonesia

by Prof. Dr. Dwia Aries Tina Pulubuhu, M.A Rector of Hasanuddin University

Panel Discussion 10.20-10.40 Morning Break and POSTER (P-1 s/d P-107) & Exhibition SESSION

10.40-12.30

Plenary II Moderator: Dr. Geni Rina Sunaryo, M.Sc

Sekretaris: Dr. Muh. Syahid Keynote Speech IV

Research and Development on High Temperature Gas Cooled Reactors and IAEA Supports to Countries Embarking on HTGRs

by Dr. Frederik Reitsma (Division of Nuclear Power IAEA) Keynote Speech V

Development of Omnidirectional Mobile Platform Using Active Dual-wheel Cater Modules and a Rocker-bogie Suspension System

by Prof. Keigo WATANABE (Okayama University, Japan)

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Keynote Speech VI ROSATOM Technology for Energy Security

by ROSATOM Keynote Speech VII

Development of Rare-Earth Element Industry by Dr. Ir. Trenggono Sutioso, MM

Keynote Speech VIII Design Development of RDE Merah Putih

by Dr. Topan Setiadipura, M.Si (BATAN) Panel Discussion

12.30-13.30 LUNCH and POSTER (P-1 s/d P-107) & Exhibition SESSION

Thur

sday

, 12 O

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PARALEL SESSION Room#1 Room#2 Room#3 Room #4

Moderator Dr. Julwan H. Purba Dr. Jalaluddin Abdul Hafid, ST, MT Dr. Muhammad Kadir

Sekretaris UNHAS Drs. Sri Kuntjoro UNHAS Jati Susilo, M.Eng

13.30-13.45 Prof. Dr. Jung Jae-cheon

KINGS Korea

Prof. Akio Gofuku OKAYAMA Univ.

Japan

O-1: Deswandri O-3: M.B. Setiawan

13.45-14.00 O-2: Sudarno O-4: Ihda Husnayani

14.00-14.15 O-5: Sigit Santoso O-7: TJ. Suryono O-29:Topan Setiadipura O-31: Julwan HP

14.15-14.30 O-6: Restu Maerani O-30: M. Subekti O-22: Mulya Juarsa O-32: Syaiful B

14.30-14.45 Coffee Break

Moderator Prof. Keigo Watanabe Drs. Sahala L. Raja Dr. Muh Syahid Drs. Amir Hamzah, M.Si.

Sekretaris Arum Puni, ST, MT UNHAS Drs. Deswandri, Meng. UNHAS

14.45-15.00 O-13: Samuel Izaak L. O-17: S. Nishihta O-21: Munawar O-25: M. As’adi

15.00-15.15 O-14: H.Yudhi Irwanto O-18: Rusdi Nur O-8: Iswadi Nur O-26: Andi Amijoyo M.

15.15-15.30 O-15: T. Imahama O-19: M. Iqbal O-23: Simon Ka’ka O-27: Kamaruddin

15.30-15.45 O-16: Rahimuddin O-20: Rafiuddin Syam O-24: Adi Tonggiroh O-28: Ruslan Buana

15.45-16.00 O-9 :Hendri Van Hoten O-12: Rafiuddin Syam O-10: Adi Maulana O-11:K. Motonaka

16.00-16.15 Closing Address

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SENTEN/ICoNETS/ISSSM Parallel Sessions ROOM#1

Moderator: Dr. Julwan H. Purba Sekretaris: UNHAS

No Time Paper Code Presentation Title

1. 13.30-13.45 - Invited Speaker NPP ENGINEERING EDUCATION IN KINGS Prof. Dr. Jung Jae-cheon (KINGS Korea)

2. 13.45-14.00

3. 14.00-14.15 O-5 OPERATOR SUPPORT SYSTEM DESIGN FOR THE OPERATION OF RSG-GAS RESEARCH REACTOR S Santoso, J Situmorang , S Bakhri, M Subekti, G.R Sunaryo

4. 14.15-14.30 O-6 V&V PLAN FOR ESF-CCS BASED FPGA USING SYSTEM ENGINEERING APPROACH Restu Maerani, Joyce Mayaka, Mohamed El Akrat, Jung Jae Cheon

5. 14.30-14.45 Coffee Break Moderator: Prof. Keigo Watanabe Sekretaris: Arum Puni, ST.,MT. 6. 14.45-15.00 O-13 THE ULTIMATE STRENGTH OF DOUBLE HULL OIL TANKER

DUE TO GROUNDING AND COLLISION Samuel Izaak Latumahina, Ganding Sitepu, Muhammad Zubair Muis Alie

7. 15.00-15.15 O-14 DEVELOPMENT OF AUTONOMOUS CONTROLLER SYSTEM OF HIGH SPEED UAV FROM SIMULATION TO READY TO FLY CONDITION Herma Yudhi Irwanto

8. 15.15-15.30 O-15 A METHOD FOR CALCULATING THE AMOUNT OF MOVEMENTS TO ESTIMATE THE SELF-POSITION OF MANTA ROBOTS Takuya Imahama, Keigo Watanabe, Kota Mikuriya and Isaku Nagai

9. 15.30-15.45 O-16 DESIGN OF OMNI DIRECTIONAL REMOTELY OPERATED VEHICLE (ROV) Rahimuddin, Hasnawiya Hasan, Haryanti A Rivai, Yanu Iskandar, Claudio P

10. 15.45-16.00 O-9 OPTIMIZATION OF PARAMETERS FOR MANUFACTURE NANOPOWDER BIOCERAMICS AT MACHINE PULVERISETTE 6 BY TAGUCHI AND ANOVA METHOD Hendri Van Hoten, Gunawarman, Ismet Hari Mulyadi, Afdhal Kurniawan Mainil dan Putra Bismantolo

10. 16.00-16.15 Closing Address

16


Moderator: Dr. Jalaluddin Sekretaris: Drs. Sri Kuntjoro


1. 13.30-13.45 - Invited Speaker COUNTER ACTION PROCEDURE GENERATION IN AN EMERGENCY SITUATION OF NUCLEAR POWER PLANTS Prof. Akio Gofuku OKAYAMA Univ.Japan

2. 13.45-14.00

3. 14.00-14.15 O-7 PRELIMINARY INVESTIGATION OF TIME REMAINING DISPLAY ON THE COMPUTER-BASED EMERGENCY OPERATING PROCEDURE T J Suryono and A Gofuku

4. 14.15-14.30 O-30 THE SIMULATOR DEVELOPMENT FOR RDE REACTOR Muhammad Subekti, Syaiful Bakhri, Geni Rina Sunaryo

5. 14.30-14.45 Coffee Break Moderator: Drs. Sahala Lumban Raja Sekretaris: UNHAS 6. 14.45-15.00 O-17 MAP GENERATION IN UNKNOWN ENVIRONMENTS BY

AUKF-SLAM USING LINE SEGMENT-TYPE AND POINT-TYPE LANDMARKS Sho Nishihta, Shoichi Maeyama, and Keigo Watanebe

7. 15.00-15.15 O-18 SUSTAINABLE MANUFACTURING BY CALCULATING THE ENERGY DEMAND DURING TURNING OF AISI 1045 STEEL Rusdi Nur, Baso Nasrullah and Asmeati

8. 15.15-15.30 O-19 THE STUDY OF PRODUCTION PERFORMANCE OF WATER HEATER MANUFACTURING BY USING SIMULATION METHOD M Iqbal, OAA Bamatraf and M Tadjuddin

9. 15.30-15.45 O-20 DESIGN MULTI-SIDES SYSTEM UNMANNED SURFACE VEHICLE (USV) ROCKET Rafiuddin Syam, Onny Sutresman, Abdullah Mappaita, Ilyas Renreng, Amiruddin, Ardi Wiranata

10. 15.45-16.00 O-12 KINEMATICS ANALYSIS OF END EFFECTOR FOR CARRIER ROBOT OF FEEDING BROILER CHICKEN SYSTEM Rafiuddin Syam1, Hairul Arsyad, Ruslan Bauna, Ilyas Renreng, Syaeful Bakhri


17


Moderator: Abdul Hafid, ST., MT. Sekretaris: UNHAS


1. 13.30-13.45 O-1 RELIABILITY ANALYSIS OF RSG-GAS PRIMARY COOLING SYSTEM TO SUPPORT AGING MANAGEMENT PROGRAM Deswandri, M.Subekti, Geni Rina Sunaryo

2. 13.45-14.00 O-2 SENSOR FAILURE DETECTION OF FASSIP SYSTEM USING PRINCIPAL COMPONENT ANALYSIS Sudarno, Mulya Juarsa, Kussigit Santosa, Deswandri, Geni Rina Sunaryo

3. 14.00-14.15 O-29 POWER PEAKING EFFECT OF OTTO FUEL SCHEME PEBBLE BED REACTOR T. Setiadipura, Suwoto, Zuhair, S. Bakhri, G.R. Sunaryo

4. 14.15-14.30 O-22 FLOW RATE AND TEMPERATURE CHARACTERISTICS IN STEADY STATE CONDITION ON FASSIP-01 LOOP DURING COMMISSIONING M Juarsa, Giarno, A. N. Rohman, G.B. Heru K., J.P. Witoko, D.T. Sony Tjahyani

5. 14.30-14.45 Coffee Break Moderator: Dr. Muh Syahid Sekretaris: Drs. Deswandri, M.Eng. 6. 14.45-15.00 O-21 THE EFFECTS OF SHIELDED METAL ARC WELDING (SMAW)

WELDING ON THE MECHANICAL CHARACTERISTICS WITH HEATING TREATMENT IN S45C STEEL Munawar, Hammada Abbas, Ahmad Yusran Aminy

7. 15.00-15.15 O-8 DESIGN OF FISHING BOAT FOR PELABUHANRATU FISHERMEN AS ONE OF EFFORT TO INCREASE PRODUCTION OF CAPTURE FISHERIES Iswadi Nur, Purwo Joko Suranto

8. 15.15-15.30 O-23 THE PNEUMATIC ACTUATORS AS VERTICAL DYNAMIC LOAD SIMULATORS ON MEDIUM WEIGHTED WHEEL SUSPENSION MECHANISM Simon Ka’ka, Syukri Himran, Ilyas Renreng and Onny Sutresman

9. 15.30-15.45 O-24 GEOLOGICAL STUDY AND REGIONAL DEVELOPMENT OF MAMBERAMO RAYA DISCTRICT OF PAPUA PROVINCE, INDONESIA Adi Tonggiroh, Asri Jaya HS, Ulva Ria Irfan

10. 15.45-16.00 O-10 STUDY ON GOLD AND BASE METAL OCCURRENCE IN ULUWAI PROSPECT, WESTERN LATIMOJONG MOUNTAIN, SOUTH SULAWESI Adi Maulana, Asri Jaya, Akira Imai


18


Moderator: Dr. Muhammad Kadir Sekretaris: Jati Susilo, M.Eng


1. 13.30-13.45 O-3 PRELIMINARY ANALYSIS OF HIGH-FLUX RSG-GAS TO TRANSMUTE AM-241 OF PWR’S SPENT FUEL IN ASIAN REGION M Budi Setiawan and S Kuntjoro

2. 13.45-14.00 O-4 KR-85M ACTIVITY AS BURNUP MEASUREMENT INDICATOR IN A PEBBLE BED REACTOR BASED ON ORIGEN2.1 COMPUTER SIMULATION I Husnayani, P M Udiyani, S Bakhri, G R Sunaryo

3. 14.00-14.15 O-31 MASTER LOGIC DIAGRAM: AN APPROACH TO IDENTIFY INITIATING EVENTS OF HTGRS J H Purba

4. 14.15-14.30 O-32 PRELIMINARY DEVELOPMENT OF ONLINE MONITORING ACOUSTIC EMISSION SYSTEM FOR THE INTEGRITY OF RESEARCH REACTOR COMPONENTS S Bakhri1, E Sumarno1, R Himawan1, T Y Akbar2, M. Subekti1, G. R. Sunaryo1

5. 14.30-14.45 Coffee Break Moderator: Drs. Amir Hamzah, MSi. Sekretaris: UNHAS 6. 14.45-15.00 O-25 ANALYZE EXPERIMENT FOR VIGAS AND PERTAMAX TO

PERFORMANCE AND EXHAUST GAS EMISSION FOR GASOLINE MOTOR 2000CC Muhamad As’adi; Diachirta Chrisna Ayu Dwiharpini Tupan

7. 15.00-15.15 O-26 COMPUTATIONAL SIMULATION ON FACIAL EXPRESSIONS AND EXPERIMENTAL TENSILE STRENGTH FOR SILICONE RUBBER AS ARTIFICIAL SKIN Andi Amijoyo Mochtar

8. 15.15-15.30 O-27 RANCANG BANGUN OMNIWHEEL ROBOT SEBAGAI SASARAN TEMBAK DINAMIS Kamaruddin, Rafiuddin Syam

9. 15.30-15.45 O-28 POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER Ruslan Bauna, Rafiuddin Syam, Hairul Arsyad, Amiruddin

10. 15.45-16.00 O-11 A SUB-TARGET APPROACH TO THE KINODYNAMIC MOTION CONTROL OF A WHEELED MOBILE ROBOT Kimiko Motonaka*, Keigo Watanabe**, and Shoichi Maeyama**


19

DAFTAR KUMPULAN JUDUL & KODE MAKALAH SENTEN-2017 No. KODE JUDUL MAKALAH DAN PENULIS PRESEN-

TASI 1. P-1 A CASE STUDY OF CONDENSING STEAM TURBINE FOR A 10 MWth

EXPERIMENTAL POWER REACTOR Sri Sudadiyo, Syaiful Bakhri, Geni Rina Sunaryo

POSTER

2. P-2 CHANNEL ANALYSIS OF OPERATION POWER FLUCTUATION FOR AP1000 REACTOR Muh. Darwis Isnaini, Deswandri, Geni Rina S.

POSTER

3. P-3 EFFECT OF F/M RATIO AGAINST NEUTRON FLUX DISTRIBUTION ON THE HTGR-10 MWth PEBBLE BED CORE Hery Adrial, Suwoto, Zuhair,Syaiful Bakhri, Geni Rina Sunaryo

POSTER

4. P-4 PRELIMANARY STUDY TO PREDICTION OF OXIDATION GRAPHITE SHELL FUEL OF HTGR ON ATWS CONDITION Elfrida Saragi, Geni Rina Sunaryo,Syaiful Bakhri

POSTER

5. P-5 DATABASE SYSTEM DEVELOPMENT FOR OPERATIONAL PARAMETER OF RSG-GAS BASED ON WEB Mike Susmikanti, Aep Saepudin, Adrian Soulisa, Muhamad Subekti, Geni Rina Sunaryo

POSTER

6. P-6 DEVELOPMENT OF ANALYSIS METHOD OF INFRARED THERMOGRAPHY FOR ELECTRICAL COMPONENT AGING MANAGEMENT Sudarno, Kussigit Santosa, Kiswanta, Deswandri, Geni Rina Sunaryo

POSTER

7. P-7 THE ON-GOING PROGRESS OF INDONESIA’S EXPERIMENTAL POWER REACTOR 10 MW AND ITS NATIONAL RESEARCH ACTIVITIES Taswanda Taryo, Rokhmadi, Syaiful Bachri, Geni Rina Sunaryo

POSTER

8. P-8 THE ANALYSIS OF THE POWER QUALITY OF THE TRANSFORMER BHT03 OF MULTIPURPOSE RESEARCH REACTOR G.A. SIWABESSY DURING THE 30 MW OPERATION Abdul Hafid, Teguh Sulistyo, Syaiful Bakhri, Geni Rina Sunaryo

ORAL

9. P-9 PENGARUH PERLAKUAN PANAS PASKA PENGELASAN TERHADAP SIFAT MEKANIK SA533-B1 SEBAGAI MATERIAL BEJANA TEKAN PWR S. Nitiswati, Andryansyah, Mudi Haryanto, Darlis, Deswandri, Geni Rina Sunaryo

POSTER

10. P-10 GAMMA DOSE RATE ANALYSIS ON 10 MWth HTGR TYPE REACTOR USING QAD-CGGP CODE Anis Rohanda, Hery Adrial, Amir Hamzah, Geni Rina Sunaryo

POSTER

11. P-11 ANALYSIS OF PRESSURE LOSS IN CHANNEL EXPERIMENTAL FACILITY Kiswanta, Sudarno, Sumantri, Deswandri, Geni Rina Sunaryo

POSTER

12. P-12 SIMULASI PERUBAHAN TEMPERATUR BERDASARKAN VARIASI DAYA DAN ALIRAN PADA HEATER KONTAK LANGSUNG UNTAI FASSIP MENGGUNAKAN SOFTWARE ChamCAD versi 6.4.1 Edy Sumarno, Mulya Juarsa, Joko P.W, Deswandri, Geni Rina S

POSTER

13. P-13 ANALISIS KEMAMPUAN PERTUKARAN KALOR TANGKI COOLER BERDASARKAN PERBEDAAN ARAH ALIRAN UNTAI HEAT SINK SYSTEM Giarno, Joko Prasetio Witoko, Mulya Juarsa, Deswandri, Geni Rina Sunaryo

POSTER

14. P-14 DEVELOPMENT OF INSTRUMENTATION AND CONTROL SYSTEM POSTER

20

AT CREEP MACHINE USING LABVIEW SOFTWARE Kussigit Santosa. Sudarno. Agus Nur Rahman, Kiswanta, Deswandri, Geni Rina Sunaryo,

15. P-15 SENSITIVITY AND UNCERTAINTY ANALYSIS OF J INTEGRAL CALCULATION OF THE RELIABILITY ASSESSMENT OF REACTOR PRESSURE VESSEL Entin Hartini, Roziq Himawan, Abdul Hafid, R.M. Subekti, Geni R. Sunaryo

POSTER

16. P-16 MODIFIKASI PRE-HEATER MENJADI PRE-COOLER PADA UNTAI UJI BETA Joko Prasetio W, Dedy H, Mulya Juarsa, Edy Sumarno, Deswandri, Geni R. Sunaryo

POSTER

17. P-17 RANCANG BANGUN ALAT PENGUKUR TEKANAN PADA UNTAI HSS FASSIP BERBASIS LABVIEW Sumantri Hatmoko, Kussigit Santosa, Agus Nur Rachman, G. Bambang Heru, Deswandri, Geni R. Sunaryo

POSTER

18. P-18 EFFECT OF CHLORIDE ION AND COPPER ION FOR TANK MATERIAL INTEGRITY OF RSG GAS PRIMARY COOLING SYSTEM Rahayu Kusumastuti, Sumaryo,Sofia Loren Butar-Butar,Sriyono, M.Subekti, Geni R. Sunaryo

POSTER

19. P-19 MONITORING TEGANGAN LDR MENGGUNAKAN ARDUINO MEGA-2560 BERBASIS LabVIEW UNTUK PENGUKURAN KERAPATAN AEROSOL G. Bambang Heru K, Alim Mardi, Joko P, Edy S, Deswandri, Geni R Sunaryo

POSTER

20. P-20 PEMBUATAN PROGRAM AKUISISI DATA PADA FASILITAS SIMULASI SISTEM PASIF (FASSIP) Agus Nur Rachman, Kussigit Santosa, Sudarno, Mulya Juarsa, Deswandri, Geni R. Sunaryo

POSTER

21. P-21 ANALISIS TERMOHIDROLIK TERAS REAKTOR TRIGA 2000 BANDUNG BERELEMEN BAKAR TIPE PELAT MENGGUNAKAN PROGRAM CFD Reinaldy Nazar, Sudjatmi KA, K. Kamajaya

POSTER

22. P-22 EFFECT OF TEMPERATURE TO ADSORPTION CAPACITY AND DISTRIBUTION COEFFICIENT ON RARE EARTH ELEMENTS ADSORPTION (Y, Dy, Gd) USING SIR Dwi Biyantoro, Agus Taftazani, Aswati Mindaryani, Supranto, Nofriady Aziz

POSTER

23. P-23 ANALISIS DISTRIBUSI TEMPERATUR KANAL TERPANAS TERAS REAKTOR TRIGA BANDUNG BERBAHAN BAKAR PELAT DENGAN PROGRAM FLUENT V.Indriati Sri Wardhani, Henky P. Rahardjo

dan Surip Widodo

POSTER

24. P-24 STUDI PERENCANAAN ENERGI KELISTRIKAN KALIMANTAN TIMUR DENGAN OPSI NUKLIR SMR Wiku Lulus Widodo, Rizki Firmansyah Setya Budi

POSTER

25. P-25 ANALISIS KEKUATAN MEKANIK ALAT BANTU ULTRASONIK UNTUK PEMERIKSAAN BEAM TUBE RSG-GAS Dedy Haryanto, Almira Citra Amelia, Aep Saepudin Catur, M. Subekti, Geni Rina Sunaryo

POSTER

26. P-26 PEMBUATAN SILIKON KARBIDA MONOLITIK DAN KARAKTERISTIKNYA Futichah, Deni Mustika, Heri Hardiyanti, Pranjono, Isfandi, Jan Setiawan

POSTER

21

27. P-27 PROYEKSI NERACA ENERGI INDONESIA HINGGA TAHUN 2060 Edwaren Liun

POSTER

28. P-28 MANAJEMEN KONSTRUKSI REAKTOR DAYA EKPERIMENTAL Mudjiono, Erlan Dewita, Dedy Priambodo

POSTER

29. P-29 EFFECT OF HEAT TREATMENT ON THE STRENGTH OF AUSTENITIC STAINLESS STEEL SS304 Andryansyah, Mudi H, Arismunandar P S, Darlis, Dwijo M, Deswandri, Geni R. Sunaryo

POSTER

30. P-30 PENENTUAN PENUMBRA PADA RADIOGRAFI BENDA BERGERAK Zaenal Abidin, Angga Fernando, Djoko Marjanto

POSTER

31. P-31 ASPEK DEMOGRAFI MENDUKUNG KEGIATAN PRA-SURVEI TAPAK PLTN DI BARELANG (BATAM, REMPANG, GALANG), KEPRI June Mellawati, Siti Alimah

POSTER

32. P-32 PERKIRAAN BIAYA EKSTERNAL DARI FASILITAS NUKLIR RDE MENGGUNAKAN SOFTWARE SIMPACT Sufiana Solihat, Wiku Lulus Widodo

POSTER

33. P-33 GAMBARAN PENERAPAN PENILAIAN DIRI DALAM PENCAPAIAN BUDAYA KESELAMATAN (STUDI KASUS DI BATAN) Farida Tusafariah, Deswandri, Arie Budianti

POSTER

34. P-34 ANALISIS POTENSI LIKUIFAKSI DI TAPAK REAKTOR DAYA EKSPERIMENTAL SERPONG Eko Rudi Iswanto, Heri Syaeful, Sriyana

POSTER

35. P-35 IDENTIFIKASI KETERDAPATAN THORIUM PADA ENDAPAN LATERIT BAUKSIT DI DAERAH NANGA TAYAP – SANDAI, KABUPATEN KETAPANG, KALIMANTAN BARAT Widodo, Putri Rahmawati, Ngadenin

POSTER

36. P-36 KAJIAN KESELAMATAN TAPAK RDE BERDASARKAN SURVEI PEDOLOGI DI KAWASAN PUSPIPTEK SERPONG, PROVINSI BANTEN Hadi Suntoko, Heni Susiati

POSTER

37. P-37 ANALISIS PENGARUH WAKTU KONSTRUKSI TERHADAP KELAYAKAN FINANSIAL PROYEK PLTN SMR DI INDONESIA DENGAN PENDEKATAN PROBABILISTIK Nuryanti, Suparman, Sufiana Solihat

POSTER

38. P-38 DETEKSI CACAT SAMPEL LAS MATERIAL SA533-B1 BEJANA TEKAN DENGAN METODA UJI TAK RUSAK Mudi Haryanto, Sri Nitiswati, Andryansyah, Deswandri, Geni R. Sunaryo

POSTER

39. P-39 ANALISIS SPASIAL TATARUANG PROGRAM RDE DI KAWASAN PUSPIPTEK, SERPONG Heni Susiati, Hadi Suntoko, Sriyana, Habib Subagio

POSTER

40. P-40 CLEARING HOUSE TEKNOLOGI NUKLIR BERBASIS STANDARDISASI SEBAGAI BASIS PELAKSANAAN CLEARING HOUSE TEKNOLOGI NUKLIR I Wayan Ngarayana

POSTER

41. P-41 STUDI DAMPAK PEMBUANGAN KONSENTRAT DESALINASI RO TERHADAP BIOTA PERAIRAN MANGGAR Siti Alimah, Heni Susiati, June Mellawati

POSTER

42. P-42 KANDUNGAN LOGAM BERAT DALAM AIR KALI PESANGGRAHAN DISEKITAR KAWASAN NUKLIR PASAR JUMAT (KNPJ) Roza Indra L, Sri Widarti, Andung Nugroho, Miki Arian S

POSTER

43. P-43 PERFORMANCE EVALUATION OF ADHOC PROTOCOLS: AODV AND DSDV FOR MOBILE NODE REQUIREMENT USING NS-2

POSTER

22

A. A. Waskita, D. Andiwijayakusuma, Deswandri, Geni R. Sunaryo 44. P-44 CONVERSSION OF CO2 TO HYDROCARBON SYNFUEL BY

UTILIZING NUCLEAR HYDROGEN COGENERATION Djati H Salimy, Syaiful Bakhri, Geni Rina Sunaryo

POSTER

45. P-45 PEMANTAUAN METEOROLOGI PADA CALON TAPAK PLTN DI DESA SEBAGIN PULAU BANGKA Denissa Beauty Syahna, Kurnia Anzhar, Slamet Suryanto

POSTER

46. P-46 DESAIN DASHBOARD UNTUK MENDUKUNG PROSES PENGAMBILAN KEPUTUSAN PEMASANGAN KAPASITOR DAYA PADA SALURAN 20 KV DI SEKITAR WILAYAH PLTN Rizki Firmansyah Setya Budi, Wiku Lulus Widodo

POSTER

47. P-47 FAKTOR PENYEBAB PENUNDAAN KONSTRUKSI PLTN DI DUNIA SEBAGAI PEMBELAJARAN UNTUK PEMBANGUNAN PLTN DI INDONESIA Dharu Dewi

POSTER

48. P-48 PEMANTAUAN GEMPA MIKRO DI CALON TAPAK PLTN MURIA JAWA TENGAH TAHUN 2015 Hajar Nimpuno Adi, Kurnia Anzhar

POSTER

49. P-49 STUDI LITERATUR “PENGUKURAN LAJU EMISI NEUTRON SECARA ABSOLUT DENGAN SISTEM MANGANESE SULPHATE BATH (MnSO4.H2O) DI CIAE, CMI, KRISS, LNE-LNHB, LNMRI, NIST, NPL, dan VNIIM”. Nazaroh

POSTER

50. P-50 KAJIAN IMPLEMENTASI PLTN DI INDONESIA: PEMBELAJARAN DARI NEGARA PENDATANG BARU Sahala Maruli Lumbanraja, Rr. Arum Puni Rijanti

POSTER

51. P-51 ANALISIS STABILITAS SISTEM KELISTRIKAN BATAM DENGAN PENAMBAHAN PEMBANGKIT LISTRIK TENAGA NUKLIR Citra Candranurani, Arief Tris Yuliyanto, Elok Satiti A, Rizki Firmansyah S.B, Rr. Arum Puni Rijanti

POSTER

52. P-52 PEMBUATAN SUMBER RADIOISOTOP 137Cs UNTUK DIGUNAKAN SEBAGAI STANDAR KALIBRASI PADA SPEKTROMETER GAMMA Aslina Br.Ginting, Yanlinastuti, Boybul, Arif Nugroho, Dian A, Gatot W,Hermawan

POSTER

53. P-53 IDENTIFIKASI ALTERASI BATUAN BERDASARKAN RASIO Th/U di TAPALANG, MAMUJU, SULAWESI BARAT I Gde Sukadana, Frederikus Dian Indrastomo, Ngadenin

POSTER

54. P-54 MOLTEN SALT REACTOR (MSR) DENGAN DAUR BAHAN BAKAR THORIUM Erlan Dewita, Sriyana

POSTER

55. P-55 KOREKSI VARIASI HARIAN UNTUK SURVEI GEOMAGNETIK DI DAERAH POTENSI URANIUM DAN THORIUM, MAMUJU SULAWESI BARAT Dwi Haryanto, Adhika Junara Karunianto

POSTER

56. P-56 ANALISA DATA GEOMAGNETIK: STUDI KASUS DI WILAYAH CALON TAPAK RDE PUSPITEK-SERPONG DAN SEKITARNYA Adhika Junara Karunianto, Dwi Haryanto, Fakhri Muhammad

POSTER

57. P-57 STUDI PENGARUH TEMPERATUR BAHAN BAKAR PADA KRITIKALITAS REAKTOR HOMOGEN (AQUEOUS HOMOGENEOUS REACTOR) MENGGUNAKAN SCALE Arif Isnaeni

POSTER

58. P-58 PEMBUATAN MIKROHIDRO UNTUK MENUNJANG KEGIATAN POSTER

23

PENELITIAN DI KAWASAN INSTALASI NUKLIR KALAN, KALBAR Slamet, Singgih Andy Nugroho,Ahmad Dayani, Eddy Soesanto

59. P-59 STUDI KETERSEDIAAN THORIUM UNTUK MENINGKATKAN KEAMANAN ENERGI NUKLIR Abimanyu Bondan, Siti Alimah, Hadi Suntoko

POSTER

60. P-60 SISTEM MANAJEMEN DOSIS PADA PROSES DAUR ULANG ZAT RADIOAKTIF TERBUNGKUS CESIUM-137 YANG SUDAH TIDAK DIGUNAKAN Suhaedi Muhammad, Rr.Djarwanti,RPS, Susyati

POSTER

61. P-61 ANALISIS DATA RADIOMETRI SEKTOR LEMAJUNG, KALAN, KALIMANTAN BARAT Heri Syaeful, Suharji, Dhatu Kamajati

POSTER

62. P-62 KARAKTERISASI HASIL IMOBILISASI ZEOLIT YANG MENGANDUNG LIMBAH THORIUM Gustri Nurliati, Yuni K. Krisnandi

POSTER

63. P-63 PENENTUAN IN-HOUSE STANDARD LOGAM TANAH JARANG Mutia Anggraini, Samin, Budi Yuli Ani, Kurnia Setiawan W

POSTER 64. P-64 ASSESSMENT OF THE RADIOLOGICAL IMPACT OF THE WASTE

TREATMENT FOR HID LAMPS CONTAINING Kr-85 AND Th-232 Moch Romli, Suhartono

POSTER

65. P-65 KARAKTERISASI LIMBAH RADIOAKTIF CAIR DAN OPTIMASI PENGOLAHAN DENGAN PENUKAR ION Ajrieh Setyawan, Ivana Oktavianita

POSTER

66. P-66 PENGAMBILAN LOGAM TANAH JARANG DALAM PASIR SENOTIM Tri Handini, Sri Sukmajaya

POSTER

67. O-27 RANCANG BANGUN OMNIWHEEL ROBOT SEBAGAI SASARAN TEMBAK DINAMIS Kamaruddin, Rafiuddin Syam

ORAL

68. O-28 POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER Ruslan Bauna, Rafiuddin Syam, Hairul Arsyad, Amiruddin

ORAL

24

Paper Code & Title listing of ICoNETS/ISSMM 2017 The 2nd International Conference on Nuclear Energy Technologies and Sciences (ICoNETS-2017)

The 4th International Symposium on Smart Material and Mechanics (ISSMM-2017) No. CODE PAPER TITLE / AUTHOR/CONTRIBUTOR PRESENTA

TION 1. P-67 SHOULD A TERRITORY BE HOME TO THE NUCLEAR WASTE

DUMP? STUDY CASE: SOUTH AUSTRALIA G. Tanjung1, S. Bakhri2, S.L. Butar-butar2

POSTER

2. P-68 AN ANALYSIS OF RADIATION PENETRATION THROUGH THE U-SHAPED CAST CONCRETE JOINTS OF CONCRETE SHIELDING IN THE MULTIPURPOSE GAMMA IRRADIATOR OF BATAN Tanti Ardiyati, Bang Rozali, Kasmudin

POSTER

3. P-69 REQUIREMENTS ANALYSIS FOR AUXILIARY POWER OF APR 1400 NPP ON SYSTEMS ENGINEERING APPROACH *M. G. Shahinoor Islam, Raouf M. Elfaramawy, Jung Jae-cheon, & Lim Hak-kyu

POSTER

4. P-70 STATIC, DYNAMIC, AND FATIGUE ANALYSIS OF THE MECHANICAL SYSTEM OF ULTRASONIC SCANNER FOR INSERVICE INSPECTION OF RESEARCH REACTORS M. Awwaluddin, Kristedjo K., Khairul Handono, Ahmad H.

POSTER

5. P-71 THE MECHATRONIC SYSTEM DESIGN OF ULTRASONIC SCANNER FOR INSERVICE INSPECTION OF RESEARCH REACTOR Khairul Handono, Kristedjo K., M. Awwaluddin and Ihsan Shobary

POSTER

6. P-72 NEUTRON DOSE RATE ANALYSIS ON HTGR-10 REACTOR USING MONTE CARLO CODE Suwoto, H. Adrial, A. Hamzah, Zuhair, S. Bakhri, G. R. Sunaryo

POSTER

7. P-73 EVALUATION OF THE AP1000 DELAYED NEUTRON PAREMETERS USING MCNP6 T.M. Sembiring, J. Susilo, S. Pinem

POSTER

8. P-74 THE CHANGE OF RADIAL POWER FACTOR DISTRIBUTION DUE TO RCCA INSERTION AT THE FIRST CYCLE CORE OF AP1000 J Susilo, L Suparlina, Deswandri, G R Sunaryo

POSTER

9. P-75 PRELIMINARY STUDY FOR ALTERNATIVE CONCEPTUAL CORE DESIGN OF THE MTR RESEARCH REACTOR Tukiran S., Surian P., Tagor MS, M. Subekti, Geni Rina Sunaryo

POSTER

10. P-76 COOLING PERFORMANCE ANALYSIS OF THE PRIMARY COOLING SYSTEM REACTOR TRIGA-2000 BANDUNG I.D. Irianto, S. Dibyo, S. Bakhri, G.R. Sunaryo

POSTER

11. P-77 ANALYSIS OF HELIUM PURIFICATION SYSTEM CAPABILITY DURING WATER INGRESS ACCIDENT IN RDE Sriyono, Rahayu Kusmastuti, Syaiful Bakhri, Geni Rina Sunaryo

POSTER

12. P-78 ANALYSIS OF RADIATION SAFETY FOR SMALL MODULAR REACTOR (SMR) ON PWR-100 MWE TYPE P. M. Udiyani, I Husnayani, Deswandri, and G. R. Sunaryo

POSTER

13. P-79 MASTER LOGIC DIAGRAM: AN APPROACH TO IDENTIFY INITIATING EVENTS OF HTGRS J H Purba

POSTER

14. P-80 MAIN STEAM LINE BREAK ACCIDENT SIMULATION OF APR1400 USING THE MODEL OF ATLAS FACILITY

POSTER

25

A S Ekariansyah, Deswandri, Geni R. Sunaryo 15. P-81 ANALYSIS ON THE ROLE OF RSG-GAS POOL COOLING SYSTEM

DURING PARTIAL LOSS OF HEAT SINK ACCIDENT Susyadi, Endiah P. H, Sukmanto D, Andi S. E, Hendro T. Syaiful B, Geni R.S.

POSTER

16. P-82 STUDY RELAP5 HELIUM PROPERTIES FOR HTGR THERMAL HYDRAULIC ANALYSIS Surip Widodo, Anis Rohanda, Muhammad Subekti, Topan Setiadipura, Syaiful Bakhri, Geni Rina S.

POSTER

17. P-83 STEADY STATE TEMPERATURE DISTRIBUTION INVESTIGATION OF HTR CORE Sudarmono, Suwoto, Syaiful Bakhri, Geni Rina Sunaryo

POSTER

18. P-84 DEVELOPMENT A COMPUTER CODES TO COUPLE PWR-GALE OUTPUT AND PC-CREAM INPUT S Kuntjoro, M Budi Setiawan, Nursinta Adi W, Deswandri, G R Sunaryo

POSTER

19. P-85 EVALUATION ON CREEP PROPERTIES OF TYPE 316SS SERIES Sri Nitiswati, Sudarno, Andryansyah, Deswandri, Geni Rina Sunaryo

POSTER 20. P-86 ANALYSIS OF JKT01 NEUTRON FLUX DETECTOR

MEASUREMENTS IN RSG-GAS REACTOR USING LabVIEW Rokhmadi, Agus Nur Rachman, Sujarwono, Taswanda Taryo, Geni Rina Sunaryo1

POSTER

21. P-87 NEUTRON FLUENCE AND DPA RATE ANALYSIS IN PEBBLE-BED HTR REACTOR VESSEL USING MCNP Amir Hamzah, Suwoto, Anis Rohanda, Hery Adrial, Syaiful Bakhri and Geni Rina Sunaryo

POSTER

22. P-88 DETERMINING COOLANT FLOW RATE DISTRIBUTION IN THE FUEL-MODIFIED TRIGA PLATE REACTOR Endiah Puji Hastuti, Surip Widodo, M. Darwis Isnaini, Geni Rina S., Syaiful B.

POSTER

23. P-89 PROBABILISTIC ANALYSIS ON THE FAILURE OF REACTIVITY CONTROL FOR THE PWR D T Sony Tjahyani, Deswandri, G R Sunaryo

POSTER

24. P-90 ULTRASONIC NON-DESTRUCTIVE PREDICTION OF SPOT WELDING SHEAR STRENGTH R. Himawan, M. Haryanto, R.M. Subekti, and G.R. Sunaryo

POSTER

25. P-91 STEADY STATE AND LOCA ANALYSIS OF KARTINI REACTOR USING RELAP5/SCDAP CODE: THE ROLE OF PASSIVE SYSTEM Anhar R. Antariksawan1, Puradwi I. Wahyono2 and Taxwim2

POSTER

26. P-92 POWER PEAKING EFFECT OF OTTO FUEL SCHEME PEBBLE BED REACTOR T. Setiadipura, Suwoto, Zuhair, S. Bakhri, G.R. Sunaryo

POSTER

27. P-93 THE EFFECT OF ZINC INJECTION ON THE INCREASING OF INCONEL 600 TT CORROSION RESISTANCES Febrianto, Sriyono, Surip Widodo, Geni Rina S

POSTER

28. P-94 NEUTRON RADIATION DAMAGE ESTIMATION IN THE CORE STRUCTURE BASE METAL OF RSG GAS S A Santa and Suwoto

POSTER

29. P-95 ANALYSIS RESPONS TO THE IMPLEMENTATION OF NUCLEAR INSTALLATIONS SAFETY CULTURE USING AHP-TOPSIS J Situmorang, I.Kuntoro, S Santoso, M Subekti, G.R Sunaryo

POSTER

30. P-96 ANALYSIS ON OPERATING PARAMETER DESIGN TO STEAM METHANE REFORMING IN HEAT APPLICATION RDE

POSTER

26

Sukmanto Dibyo, Geni Rina Sunaryo, Syaiful Bakhri, Zuhair, Ign.Djoko Irianto

31. P-97 RADIOLOGICAL RISK ESTIMATION OF 137CESIUM ON A NEAR SURFACE DISPOSAL FACILITY BY USING RESRAD ONSITE APPLICATION Budi Setiawan1*, Suci Prihastuti2, Setyo S Moersidik3

POSTER

32. P-98 THE SIMULATOR DEVELOPMENT FOR RDE REACTOR Muhammad Subekti, Syaiful Bakhri, Geni Rina Sunaryo

POSTER

33. P-99 MODELLING THE RADIOLYSIS OF RSG-GAS PRIMARY COOLING WATER: PRELIMINARY STUDY Sofia Loren Butarbutar, Rahayu Kusumastuti, M. Subekti, Geni Rina Sunaryo

POSTER

34. P-100 INFLUENCE OF SOL CONCENTRATION TO PARTICLE DIAMETER OF CERIUM STABILIZED ZIRCONIUM MADE BY EXTERNAL GELATION Sukarsono, Meniek Rahmawati, Sri Rinanti S, Dedy Husnurrofiq, Kristanti And Ariyani Dewi K

POSTER

35. P-101 REACTIVITY COEFFICIENT CALCULATION FOR AP1000 REACTOR USING THE NODAL3 CODE Surian Pinem, Tagor Malem Sembiring, Deswandri, Geni Rina Sunaryo

POSTER

36. P-102 STUDY ON CHARACTERISTIC OF TEMPERATURE COEFFICIENT OF REACTIVITY FOR PLUTONIUM CORE OF PEBBLED BED REACTOR Zuhair, Suwoto, T. Setiadipura, S. Bakhri, G.R. Sunaryo

POSTER

37. P-103 EFFECTS OF THE APPLICATION OF THE NEW NUCLEAR DATA LIBRARY ENDF/B TO THE CRITICALITY ANALYSIS OF AP1000 Iman Kuntoro,T.M. Sembiring, Deswandri, G.R. Sunaryo

POSTER

38. P-104 A PRELIMINARY DESIGN OF APPLICATION OF WIRELESS IDENTIFICATION AND SENSING PLATFORM ON EXTERNAL BEAM RADIOTHERAPY Heranudin1,2 and S. Bakhri3

POSTER

39. P-105 LEACHING KINETIC OF Nd. Y, Pr AND Sm IN RARE EARTH HYDROXIDE (REOH) USE NITRIC ACID MV Purwani and Suyanti

POSTER

40. P-106 PRELIMINARY DEVELOPMENT OF ONLINE MONITORING ACOUSTIC EMISSION SYSTEM FOR THE INTEGRITY OF RESEARCH REACTOR COMPONENTS S Bakhri1, E Sumarno1, R Himawan1, T Y Akbar2, M. Subekti1, G. R. Sunaryo1

POSTER

41. P-107 SOLVENT SELECTION FOR EXTRACTION OF NEODYMIUM CONCENTRATES OF MONAZITE SAND PROCESSED PRODUCT Moch Setyadji. MV Purwani

POSTER

42. - COUNTER ACTION PROCEDURE GENERATION IN AN EMERGENCY SITUATION OF NUCLEAR POWER PLANTS Akio Gofuku (OKAYAMA University Japan)

Invited Speaker

43. - NPP ENGINEERING EDUCATION IN KINGS

Jung Jae-cheon (KINGS Korea) Invited

Speaker 44. O-1 RELIABILITY ANALYSIS OF RSG-GAS PRIMARY COOLING

SYSTEM TO SUPPORT AGING MANAGEMENT PROGRAM Deswandri, M.Subekti, Geni Rina Sunaryo

ORAL

45. O-2 SENSOR FAILURE DETECTION OF FASSIP SYSTEM USING PRINCIPAL COMPONENT ANALYSIS

ORAL

27

Sudarno, Mulya Juarsa, Kussigit Santosa, Deswandri, Geni Rina Sunaryo

46. O-3 PRELIMINARY ANALYSIS OF HIGH-FLUX RSG-GAS TO TRANSMUTE AM-241 OF PWR’S SPENT FUEL IN ASIAN REGION M Budi Setiawan and S Kuntjoro

ORAL

47. O-4 KR-85M ACTIVITY AS BURNUP MEASUREMENT INDICATOR IN A PEBBLE BED REACTOR BASED ON ORIGEN2.1 COMPUTER SIMULATION I Husnayani, P M Udiyani, S Bakhri, G R Sunaryo

ORAL

48. O-5 OPERATOR SUPPORT SYSTEM DESIGN FOR THE OPERATION OF RSG-GAS RESEARCH REACTOR S Santoso, J Situmorang , S Bakhri, M Subekti, G.R Sunaryo

ORAL

49. O-6 V&V PLAN FOR ESF-CCS BASED FPGA USING SYSTEM ENGINEERING APPROACH Restu Maerani, Joyce Mayaka, Mohamed El Akrat, Jung Jae Cheon

ORAL

50. O-7 PRELIMINARY INVESTIGATION OF TIME REMAINING DISPLAY ON THE COMPUTER-BASED EMERGENCY OPERATING PROCEDURE T J Suryono and A Gofuku

ORAL

51. O-8 DESIGN OF FISHING BOAT FOR PELABUHANRATU FISHERMEN AS ONE OF EFFORT TO INCREASE PRODUCTION OF CAPTURE FISHERIES Iswadi Nur, Purwo Joko Suranto

ORAL

52. O-9 OPTIMIZATION OF PARAMETERS FOR MANUFACTURE NANOPOWDER BIOCERAMICS AT MACHINE PULVERISETTE 6 BY TAGUCHI AND ANOVA METHOD Hendri Van Hotena, Gunawarmanb, Ismet Hari Mulyadib, Afdhal Kurniawan Mainila dan Putra Bismantoloa

ORAL

53. O-10 STUDY ON GOLD AND BASE METAL OCCURRENCE IN ULUWAI PROSPECT, WESTERN LATIMOJONG MOUNTAIN, SOUTH SULAWESI Adi Maulana1, Asri Jaya1, Akira Imai2

ORAL

54. O-11 A SUB-TARGET APPROACH TO THE KINODYNAMIC MOTION CONTROL OF A WHEELED MOBILE ROBOT Kimiko Motonaka*, Keigo Watanabe**, and Shoichi Maeyama**

ORAL

55. O-12 KINEMATICS ANALYSIS OF END EFFECTOR FOR CARRIER ROBOT OF FEEDING BROILER CHICKEN SYSTEM Rafiuddin Syam1, Hairul Arsyad1, Ruslan Bauna1, Ilyas Renreng1, Syaeful Bakhri2

ORAL

56. O-13 THE ULTIMATE STRENGTH OF DOUBLE HULL OIL TANKER DUE TO GROUNDING AND COLLISION Samuel Izaak Latumahina, Ganding Sitepu, Muhammad Zubair Muis Alie

ORAL

57. O-14 DEVELOPMENT OF AUTONOMOUS CONTROLLER SYSTEM OF HIGH SPEED UAV FROM SIMULATION TO READY TO FLY CONDITION Herma Yudhi Irwanto

ORAL

58. O-15 A METHOD FOR CALCULATING THE AMOUNT OF MOVEMENTS TO ESTIMATE THE SELF-POSITION OF MANTA ROBOTS Takuya Imahama, Keigo Watanabe, Kota Mikuriya and Isaku Nagai

ORAL

59. O-16 DESIGN OF OMNI DIRECTIONAL REMOTELY OPERATED VEHICLE (ROV) Rahimuddin1, Hasnawiya Hasan2, Haryanti A Rivai1, Yanu Iskandar3,

ORAL

28

Claudio P3 60. O-17 MAP GENERATION IN UNKNOWN ENVIRONMENTS BY AUKF-

SLAM USING LINE SEGMENT-TYPE AND POINT-TYPE LANDMARKS Sho Nishihta, Shoichi Maeyama, and Keigo Watanebe

ORAL

61. O-18 SUSTAINABLE MANUFACTURING BY CALCULATING THE ENERGY DEMAND DURING TURNING OF AISI 1045 STEEL Rusdi Nur, Baso Nasrullah and Asmeati

ORAL

62. O-19 THE STUDY OF PRODUCTION PERFORMANCE OF WATER HEATER MANUFACTURING BY USING SIMULATION METHOD M Iqbal1, OAA Bamatraf2 and M Tadjuddin3

ORAL

63. O-20 DESIGN MULTI-SIDES SYSTEM UNMANNED SURFACE VEHICLE (USV) ROCKET Rafiuddin Syam, Onny Sutresman, Abdullah Mappaita, Ilyas Renreng, Amiruddin, Ardi Wiranata

ORAL

64. O-21 THE EFFECTS OF SHIELDED METAL ARC WELDING (SMAW) WELDING ON THE MECHANICAL CHARACTERISTICS WITH HEATING TREATMENT IN S45C STEEL Munawar1, Hammada Abbas2, Ahmad Yusran Aminy2

ORAL

65. O-22 FLOW RATE AND TEMPERATURE CHARACTERISTICS IN STEADY STATE CONDITION ON FASSIP-01 LOOP DURING COMMISSIONING M Juarsa, Giarno, A. N. Rohman, G.B. Heru K., J.P. Witoko, D.T. Sony Tjahyani

ORAL

66. O-23 THE PNEUMATIC ACTUATORS AS VERTICAL DYNAMIC LOAD SIMULATORS ON MEDIUM WEIGHTED WHEEL SUSPENSION MECHANISM Simon Ka’ka1, Syukri Himran2, Ilyas Renreng2 and Onny Sutresman2

ORAL

67. O-24 GEOLOGICAL STUDY AND REGIONAL DEVELOPMENT OF MAMBERAMO RAYA DISCTRICT OF PAPUA PROVINCE, INDONESIA Adi Tonggiroh, Asri Jaya HS, Ulva Ria Irfan

ORAL

68. O-25 ANALYZE EXPERIMENT FOR VIGAS AND PERTAMAX TO PERFORMANCE AND EXHAUST GAS EMISSION FOR GASOLINE MOTOR 2000CC Muhamad As’adi; Diachirta Chrisna Ayu Dwiharpini Tupan

ORAL

69. O-26 COMPUTATIONAL SIMULATION ON FACIAL EXPRESSIONS AND EXPERIMENTAL TENSILE STRENGTH FOR SILICONE RUBBER AS ARTIFICIAL SKIN Andi Amijoyo Mochtar

ORAL

70. O-29 POWER PEAKING EFFECT OF OTTO FUEL SCHEME PEBBLE BED REACTOR T. Setiadipura, Suwoto, Zuhair, S. Bakhri, G.R. Sunaryo

ORAL

71. O-30 THE SIMULATOR DEVELOPMENT FOR RDE REACTOR Muhammad Subekti, Syaiful Bakhri, Geni Rina Sunaryo

ORAL

72. O-31 MASTER LOGIC DIAGRAM: AN APPROACH TO IDENTIFY INITIATING EVENTS OF HTGRS J H Purba

ORAL

73. O-32 PRELIMINARY DEVELOPMENT OF ONLINE MONITORING ACOUSTIC EMISSION SYSTEM FOR THE INTEGRITY OF RESEARCH REACTOR COMPONENTS S Bakhri1, E Sumarno1, R Himawan1, T Y Akbar2, M. Subekti1, G. R. Sunaryo1

ORAL

29

2017

30

P-1 THE CONDENSING STEAM TURBINE CASE STUDY FOR A 10 MWth EXPERIMENTAL POWER REACTOR

Sri Sudadiyo, Syaiful Bakhri, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety, BATAN Kawasan PUSPIPTEK Building 80, Serpong, Tangerang 15310

Telp/Fax: (021)7560012/(021)7560913, E-mail : [email protected]

ABSTRAK STUDI KASUS TURBIN UAP KONDENSASI UNTUK REAKTOR DAYA EKSPERIMENTAL 10 MWth. Sistem pendingin Reaktor Daya Eksperimental (RDE) 10 MWth mengintegrasikan siklus blower helium sebagai sirkuit primer dan siklus turbin uap sebagai sirkuit sekunder. Sistem ini dapat mencapai effisiensi termal dan tenaga mekanis yang lebih tinggi melalui pemanfaatan yang tepat dari energi dengan meminimalkan kehilangan energi menuju minimum. Dalam penelitian ini, pengaruh beban operasional turbin uap kondensasi seperti tekanan uap, tekanan kondensor, dan temperatur masuk turbin terhadap keluaran daya dan efisiensi termal RDE diinvestigasi. Hasil studi ini dapat dimanfaatkan untuk menyederhanakan desain konsep RDE dengan nilai efisiensi dan tenaga lebih sesuai. Simulasi Cycle-Tempo telah dilakukan untuk meneliti pengaruh parameter yang disebutkan di atas pada siklus turbin uap. Berbagai komponen dari sirkuit primer dan sekunder dimodelkan termasuk generator uap, blower, pompa, kondensor, turbin dan generator. Pada makalah ini, jenis turbin SST-60, SST-100, SST-111, SST-300, dan SST-600 digunakan untuk simulasi. Metode konservasi digunakan untuk menyelesaikan persamaan massa, momentum dan energi untuk memperoleh sifat aliran helium dalam siklus blower dan air/uap dalam siklus turbin. Untuk kasus SST-100, hasil simulasi memberikan nilai efisiensi turbin 92,26 %, efisiensi termal optimum 25,5 %, dan tenaga mekanis 3650,52 kW pada putaran blower 3255 rpm. Karakteristika komponen dari generator uap, kondensor, dan pompa juga ditunjukkan dengan menampilkan hasil yang baik untuk kinerja yang dicapai. Kata kunci: RDE 10 MWth, siklus Rankine, turbin uap, kondensor, efisiensi

P-2 CHANNEL ANALYSIS OF OPERATION POWER FLUCTUATION FOR AP1000 REACTOR

Muh. Darwis Isnaini, Deswandri, Geni Rina S. Center for Nuclear Reactor Technology and Safety, BATAN,

PUSPIPTEK Area Building no.80 Serpong, Tangerang Selatan, 15310 Indonesia email: [email protected]

ABSTRACT

CHANNEL ANALYSIS OF OPERATION POWER FLUCTUATION FOR AP1000 REACTOR. A study to analyze the influence of operation power fluctuations on channel analysis of AP1000 reactor has been carried out. The calculation was conducted as channel analysis, because sub-channel analysis could not be done by using COBRA-EN code for transient or power as time function. The calculations found that hot channel at peak linear power of 42.66 kW/m could represent hot sub-channel with peak linear power of 48.88 kW/m, fairly well. Between the two models, it was found that the peak center line fuel temperature, the peak radial average fuel temperature and MDNBR were 5.54%, -2.83% and 6.35%, respectively. For calculations of operation power fluctuations the 12-3-3-3-0.5-0.5-0.5-0.5-1 mode was used as a model for the hot channel using fuel and clad thermal properties as temperature functions. The results showed that on full power, 50-percent power and 110-percent power, the peak center line fuel temperatures were found to be 1697.25°C, 840.25°C and 1868.75°C, respectively. Moreover, the MDNBR were 2.65, 5.32 and 2.41, respectively. It was concluded that the reactor could be operated safely, while operation power fluctuation occurred. Keyword: channel analysis, operation power fluctuation, COBRA-EN, AP1000.

31

P-3 EFFECT OF F/M RATIO AGAINST NEUTRON FLUX DISTRIBUTION ON THE HTGR-10 MWth PEBBLE BED CORE

Hery Adrial, Zuhair,Suwoto, Syaiful Bakhri, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety

Kawasan Puspiptek Gd.80, Serpong, Tangerang Selatan Email: hery-adr @batan.go.id

ABSTRACT EFFECT OF F / M RATIO AGAINST OF NEUTRON FLUX DISTRIBUTION ON THE HTGR-10 MWTH PEBBLE BED CORE. The research on the effect of F/M ratio on neutron flux distribution in the HTGR-10 MWth pebble bed core has been done. The goal of this research is to know the effect caused by variations of pebble fuel ratio with pebble moderator (F/M) on neutron flux distribution at HTGR-10 MWth core and to obtain optimal F/M ratio to be applied in the HTGR- 10 MWth core. In this research, the first step is completed modeling of HTGR-10 MWt then calculation of the neutron energy spectrum and neutron flux calculation, with variations F/M = 100: 0, 80:20, 60:40, 57; 43, 52; 48, 50 : 50, 40.60, and 20:80. Modeling and calculation are performed using the program package EGS99304 with 36 group vitamin C, MCNP6.1 and VisEd. The calculation results have shown that the resulting neutron spectrum is identical to the neutron spectrum that occurs on the core of the nuclear reactor. The highest neutron flux distribution of 1.32E + 14 n/cm2 sec occurred at F: M = 40:60 with the position at the center of the reactor core. The fuel ratio is also in the range of 40% to 60% to apply to the HTGR10 MWth core with multiplication factor value is in the range of 1.08 to 1.16. From the results of calculation, it can be concluded that highest flux the HTGR-10 MWth is 1.32 E+14 n/cm2 sec with F/M ratio is 40:60, while the variation F/M ratio which can be used on HTGR-10 MWth is from the ratio of 40:60 to 60:40. Keywords: neutron flux, neutron spectrum, HTGR-10 MWth, pebble bed core, F/M ratio

P-4 PRELIMANARY STUDY TO PREDICTION OF OXIDATION GRAPHITE SHELL FUEL OF HTGR ON ATWS CONDITION

Elfrida Saragi1, Geni Rina Sunaryo2,Syaiful Bakhri3

1,2,3Centre for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Puspiptek Area, Building 80, Serpong,15310, Indonesia.

Email: [email protected] .

ABSTRACT PRELIMANARY STUDY TO PREDICTION OF OXIDATION GRAPHITE SHELL FUEL OF HTGR ON ATWS CONDITION. One form of fuel HTGR is a sphere shape. Spherical fuel of high temperature gas-cooled reactor (HTGR) is coated in graphite. One of the causes of the weakness of graphite structure is due to graphite degradation. One of the accidents occurring in a high temperature gas cooled reactor (HTGR) type is air in-leakage to the primary system called air ingress. Air ingress is preceded by a pressure drop (D-LOFC) that leads to the degradation of the graphite shell fuel due to chemical reactions between oxygen and graphite at temperatures above 950 0C and heating reactor core. The events of air ingress occurs in conditions of D-LOFC on the anticipated transient without scram (ATWS). The air ingress is considered as hypothetical scenario. The purpose of this study is to find out the strength of the structure of the graphite shell of HTGR due to oxidation occurring on air ingress-D-LOFC conditions. To determine the effect of oxidation on the integrity of the fuel shell structure, the rate of oxidation is estimated using computational simulations with GRSAC Code. The computational simulation used the data PBMR 400 MWt. The computational simulation is resulted in an oxidation rate of 300 g/min for 60 hours with an ATWS delayed of 2,000 minutes with depressurization for 50 minutes. Fractional weight loss due to oxidation of 0.49 with a time of 125 hours is obviously smaller than the Idaho National Laboratory (INL) results of 0.647. Therefore, strength of the mechanical structure of graphite fuel shell is still in good condition. Keywords: Shell Graphite, Oxidation rate, Mechanical strength ,ATWS, GRSAC code.

32

P-5 DATABASE SYSTEM DEVELOPMENT FOR OPERATIONAL PARAMETER OF RSG-GAS BASED ON WEB

Mike Susmikanti1, Aep Saepudin2, Adrian Soulisa3, Muhamad Subekti1, Geni Rina1

1Center for Nuclear Reactor Technology and Safety, 2Center for Multipurpose Reactor

National Nuclear Energy Agency of Indonesia (BATAN), PUSPIPTEK Area Building 80 Serpong, Tangerang Selatan 15310, Indonesia

3Faculty of Computer Science, Gunadarma University Margonda, Depok, Indonesia

Email: [email protected]

ABSTRACT DATABASE SYSTEM DEVELOPMENT FOR OPERATION PARAMETER OF RSG-GAS BASED ON WEB. The Information and data collection for parameter operation are important well documented for the aging management of research reactor RSG-GAS. The RSG-GAS multipurpose reactor has been operating for almost 30 years so it is necessary to be evaluated on using the database system for the management operation of RSG-GAS especially ways of working the structure, system, and components (SSC). The system database SSC of RSG-GAS still static. It is not easy to find the necessary data. The system database based on WEB are expected to be used in an online system to obtain information on operation parameter of each component of the system. The purpose of this study is to create and develop RSG-GAS database system for parameter operation by utilizing web-based technologies. The system database are expected contain the data and information of parameter operation which can be integrated with some certain user and administration to obtain the information of operation processing RSG-GAS. By using this system database can be monitor, that the operation still within the operating limit. The system database has made for The operation recording of some components. The graph of operation has made for some component. The database system based on WEB has built using bootstrap framework technology, PHPMySQL. Keyword: Database System, RSG-GAS, Operation parameter, WEB, PHPMySQL

P-6 DEVELOPMENT OF ANALYSIS METHOD OF INFRARED THERMOGRAPHY FOR ELECTRICAL COMPONENT AGING MANAGEMENT

Sudarno1, Kussigit Santosa1, Kiswanta1, Deswandri1, Geni Rina Sunaryo1

1Center for Nuclear Reactor Technology and Safety - BATAN Puspiptek Area, Building no. 80 - Serpong, South Tangerang, Indonesia

Email : [email protected]

ABSTRACT DEVELOPMENT OF ANALYSIS METHOD OF INFRARED THERMOGRAPHY FOR ELECTRICAL COMPONENT AGING MANAGEMENT. The GA Siwabessy multiple purpose reactor (GAS MPR) is a research reactor that has been operating for more than twenty years. In the IAEA Safety Standard Series (DS-272) on Safety Requirements of Research Reactors, it is indicated that the sufficient action / effort must be performed for testing and observation in order to detect, evaluate and mitigate the effects of aging. In this research, the development of inspection methods of electrical systems using infrared thermography has been done. The objective of this research is to propose an image processing method to infrared thermography inspection results, so that it simplify the hotspot extraction of infrared thermography inspection results. The methodology of this research is to apply image processing algorithms such as Independent Component Analysis (ICA) for image segmentation and k-means method clustering. The program uses input of infrared thermal image and provides outputs of featured images as segmentation and clustering results.The test results indicated that both methods can be used to improve hotspot detection of the infrared thermography inspection results. Interpretation of clustering result is easier than segmentation. However, segmentation test result with fastICA provided more detail information than clustering method. Keywords: Infrared Thermography, Independent Component Analysis (ICA), segmentation, k-means clustering.

33

P-7 THE ON-GOING PROGRESS OF INDONESIA’S EXPERIMENTAL POWER REACTOR 10 MW AND ITS NATIONAL RESEARCH ACTIVITIES

Taswanda Taryo, Rokhmadi, Syaiful Bakhri, Geni Rina Sunaryo

Center for Technology and Nuclear Reactor Safety-BATAN, Kawasan Puspiptek, Gedung 80, Serpong 15310, INDONESIA

Email address: [email protected]

ABSTRACT THE ON-GOING PROGRESS OF INDONESIA’S EXPERIMENTAL POWER REACTOR 10 MW (RDE-10) AND NATIONAL RESEARCH ACTIVITIES.The RDE-10 program was firstly launched to the Agency for National Development Planning (BAPPENAS) in 2014. The RDE-10 program is expected to devote positive impacts to community prosperity, self-reliance and sovereignty of Indonesia. The RDE-10 availability will be able to accelerate advance nuclear technology development and hence Indonesia will turnout to be the nuclear champion in the ASEAN region. The application of RDE-10 performs an added value of local content and hence develops a model to fulfill the need of raw-material industry which is still imported from other countries. The success of development, operation, maintenance and utilization of RDE-10 will be able to enhance public acceptance on nuclear technology in Indonesia. This paper entitles background and design specification of RDE-10, challenging on budgeting and site licensing, national capacity building and its significant implementing research activities, concluding remarks, acknowledgement and references. Indeed, the paper can be assigned as a reference in planning, construction and operation of RDE-10 in Indonesia. Keywords: RDE-10, Indonesia, current status, development.

P-8 THE ANALYSIS OF THE POWER QUALITY OF THE TRANSFORMER BHT03 OF MULTIPURPOSE RESEARCH REACTOR G.A. SIWABESSY DURING THE 30 MW

OPERATION

Abdul Hafid1, Teguh Sulistyo2, Syaiful Bakhri1, Geni Rina Sunaryo1 1 Center For Nuclear Reactor Technology and Safety BATAN, PUSPIPTEK AREA Building No. 80, Tangsel

2 Center for Multipurpose Reactor BATAN, PUSPIPTEK AREA Building No. 30, Tangerang Selatan email: [email protected]

ABSTRAK

ANALISIS KUALITAS DAYA TRANSFORMATOR LISTRIK BHT03 REAKTOR SERBA GUNA G.A. SIWABESSY SAAT OPERASI REAKTOR 30 MW. Usia Reaktor Serba Guna G. A. Siwabessy (RSG GAS) sudah 30 tahun. Penuaan merupakan masalah yang harus dikontrol. Pasokan listrik pada komponen senantiasa diharapkan memiliki kualitas yang baik. Listrik RSG-GAS dipasok dari Perusahaan Listrik Negara melalui transformator, salah satunya adalah transformator BHT03. Telah dilakukan pengukuran pada sisi tegangan rendah transformator BHT03 RSG-GAS. Pengukuran dilakukan guna mendapatkan data kelayakan daya listrik RSG GAS. Tujuan penelitian ini adalah untuk mengetahui kulitas daya listrik transformator BHT03 RSG-GAS pada saat operasi reaktor 30 MW. Metode yang digunakan adalah pengukuran langsung dengan menggunakan alat ukur Power Quality Analyzer (PQA) 3197. Hasil analisis dari pengukuran menunjukkan bahwa pada pengoperasian 30 MW ini, terdapat aliran arus pada fasa netral sebesar 61 A. Dari hasil analisis perhitungan ketidakseimbangan arus listrik pada transformator BHT03 sebesar 1,98%. Faktor daya listrik sangat baik dengan nilai terendah 0,92 dan tertingi 0,95, rata-rata sebesar 0,94. Kualitas daya listrik pada transformator BHT03 memenuhi syarat karena Persyaratan Umum Instalasi Listrik (PUIL) menyatakan bahwa ketidak seimbangan listrik adalah kurang dari 20%. Kata kunci: kualitas daya, transformator, operasi 30 MW, arus listrik, ketidakseimbangan

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P-9 PENGARUH PERLAKUAN PANAS PASCA PENGELASAN TERHADAP SIFAT MEKANIK SA533-B1 SEBAGAI MATERIAL BEJANA TEKAN PWR

S. Nitiswati, Andryansyah, Mudi Haryanto, Darlis, Deswandri, Geni Rina Sunaryo

Pusat Teknologi Dan Keselamatan Reaktor Nuklir – Badan Tenaga Nuklir Nasional Puspiptek Gedung No. 80, Tangerang Selatan 15313,


ABSTRAK PENGARUH PERLAKUAN PANAS PASCA PENGELASAN TERHADAP SIFAT MEKANIK SA533-B1 SEBAGAI MATERIAL BEJANA TEKAN PWR. Bejana tekan reaktor air bertekanan (Pressurized Water Reactor) adalah komponen utama paling kritis di PLTN yang dikonstruksi dengan cara di las circumferential dan longitudinal. Dalam proses pengelasan akan menimbulkan tegangan sisa yang merupakan tegangan internal material dan mempunyai potensi menurunkan sifat mekanik material. Ada cara untuk mengurangi tegangan sisa akibat pengelasan yaitu dengan memberikan perlakuan panas. Penelitian ini mendiskusikan pengaruh perlakuan panas pasca pengelasan terhadap sifat mekanik SA 533-B1 sebagai material bejana tekan PWR. Tujuan penelitian ini adalah untuk menginvestigasi sifat mekanik SA533-B1 yang telah diberikan perlakuan panas pada temperatur 400°C dengan variasi holding time dari 3 jam, 50 jam, 75 jam dan 100 jam. Metode yang digunakan adalah dengan melakukan pengujian mekanik terdiri dari pengukuran kekerasan dan pengujian tarik logam las-lasan SA533-B1 pada posisi cross weld. Model bidang patahan dari hasil pengujian tarik juga dipelajari. Disimpulkan bahwa material SA533-B1 yang mengalami perlakuan panas pada temperatur 400°C dengan holding time 75 jam menunjukkan nilai kekerasan dan sifat mekanik pada kondisi yang hampir sama dengan material segar atau terjadi recovery mendekati ke keadaan semula. Kata kunci: Perlakuan panas, SA533-B1, sifat mekanik, bejana tekan PWR.

P-10 GAMMA DOSE RATE ANALYSIS ON 10 MWth HTGR TYPE REACTOR USING QAD-CGGP CODE

Anis Rohanda1, Hery Adrial1, Amir Hamzah1, Syaiful Bakhri1, GR Sunaryo1 1 Center for Nuclear Reactor Technology and Safety (PTKRN) - BATAN

Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 Email: [email protected]

ABSTRACT

GAMMA DOSE RATE ANALYSIS on 10 MWth HTGR TYPE REACTOR USING QAD-CGGP CODE. Experimental Power Reactor (RDE) is one of BATAN's flagship programs to encourage the government to accelerate the implementation of nuclear in the national energy system mix. By demonstrating the operation of the RDE, it is expected to increase the level of public acceptance and to demonstrate the readiness of human resources in power reactor operation. Design base of RDE is high temperature gas cooled reactor (HTGR) small scale with about 10 MW of thermal power. This activity needs to be supported through various studies related to design and safety aspects. One of the most important forms of study is the analysis of gamma dose rates. Gamma dose rate reflects the penetration of gamma radiation on various media. Gamma dose rate information is very useful for the design of the radiation shield and as a basis for organizing working hours for radiation workers. The methodology were compiling complete data specifications of core and building of HTGR 10 MWth and modeling the reactor building by using EASYQAD which is a Graphical User Interface (GUI) for QAD-CGGP code used for gamma dose rate analysis. Strong gamma source (source term) is determined by using code ORIGEN-ARP module SCALE program version 5. The results of the analysis show that gamma dose rates distribution are ranging from 2.08 to 2.06 × 105 μSv/h. Accordly PWR type reactor design restrictions issued by the US-NRC then the reactor area is categorized in zone I, III, IV and V. Areas with low radiation levels (zone I) with unlimited number of workers are obtained dose rates of <2.5 ×101 μSv/h, whereas areas with high radiation levels dose rate of 1.0 × 104 - 1.0 × 105 μSv/h with access to the number of workers is very limited. Keywords: gamma dose, HTGR, RDE, QAD-CGGP

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P-11 ANALYSIS OF PRESSURE LOSS IN CHANNEL EXPERIMENTAL FACILITY

Kiswanta, Sudarno, Deswandri, Geni Rina Sunaryo Center for Nuclear Reactor Technology and Safety (CNRTS)-BATAN

Gd. 80 Kawasan Puspiptek-Serpong, Tangerang Selatan, Banten Telp.021-7560912 / Fax. 021-7560913


ABSTRACT ANALYSIS OF PRESSURE LOSS IN CHANNEL EXPERIMENTAL FACILITY. Channel Experiment Facility (ExNal) is an experimentation facility that can be used to simulate the flow rate of coolant to the research reactor. ExNal can be used to understand the potential occurrence of vibration in the fuel plate. Potential vibration is necessary to know due to vibration can accelerate materials fatique and ultimately could lead to material damage such as cracked fuel and even rupture. From previous research facilities have not been able to achieve the critical speed of 18.062 m/s. The purpose of this study is to analyze the loss of pressure that occurs in parts ExNal. That methodology of research done by calculating the pressure loss through the approach of critical velocity coolant flow that can be known pump efficiency at the ExNal facility. The result showed 0.64 bar of total loss pressure and 64% of pump efficiency. Therefore, parameters such as fluid velocity, flow rate and pressure has not approached the expected results, the capacity of the pump must be improved. Keywords: experimental channel, the pressure loss, pump efficiency, critical speed, vibration

P-12 SIMULASI PERUBAHAN TEMPERATUR BERDASARKAN VARIASI DAYA DAN ALIRAN PADA HEATER KONTAK LANGSUNG UNTAI FASSIP MENGGUNAKAN SOFTWARE

ChamCAD versi 6.4.1

Edy Sumarno, Mulya Juarsa, Joko P.W, Deswandri, Geni Rina S Pusat Teknologi dan Keselamatan Reaktor Nuklir -BATAN,

Kawasan Puspiptek Gd. 80 Serpong Tangerang Selatan Banten Telp.021-7560912/Fax.021-7560913 Kode pos 1315


ABSTRAK Untai FASSIP merupakan fasilitas eksperimen dimana aliran fluida pada sistem tersebut bergerak dengan sendiri karena adanya perbedaan temperatur pada dua sisi yang berbeda. Sirkulasi alami merupakan kemampuan fluida untuk bersikulasi secara berkesinambungan yang disebabkan oleh perbedaan kerapatan densitas fluida karena adanya beda temperatur. Salah satu komponen utama untai FASSIP adalah digunakannya heater sebagai sumber panas. Untuk mendapatkan panas yang sesuai dengan yang dibutuhkan pada untai FASSIP memerlukan heater sebagai sumber panasnya. Tujuan penelitian adalah mendesain dan menghitung serta menentukan besaran diameter kawat kanthal yang digunakan agar kebutuhan panas dan capaian temperatur yang diinginkan dapat dipenuhi. Metode yang digunakan adalah dengan cara melakukan perhitungan besaran resistan dan panjang kawat kanthal, dengan cara menentukan terlebih dahulu besaran diameter dan daya yang digunakan pada kawat kanthal tersebut. Penentuan besaran temperatur dilakukan dengan cara memvariasikan daya dan aliran menggunakan program ChamCAD versi 6.4.1. Hasil desain untuk heater kontak lansung yang menggunakan kawat kanthal dengan diameter 2 mm dan daya maksimum sebesar 5 kW, didapat panjang kanthal sepanjang 17,623 meter. Hasil simulasi dengan daya 1 kW didapat flowrate 0,1 liter/menit sebesar 113,34oC dan pada flowrate 1,0 liter/menit sebesar 41,33oC. Kata kunci: Heater, Untai FASSIP, Software ChamCAD 6.4.1

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P-13 ANALISIS KEMAMPUAN PERTUKARAN KALOR TANGKI COOLER BERDASARKAN PERBEDAAN ARAH ALIRAN UNTAI HEAT SINK SYSTEM

Giarno, Joko Prasetio Witoko, Mulya Juarsa, Deswandri, Geni Rina Sunaryo

PTKRN-BATAN, Kawasan Puspiptek Serpong, Gedung 80,15310 email: [email protected]

ABSTRAK

ANALISIS KEMAMPUAN PERTUKARAN KALOR TANGKI COOLER BERDASARKAN PERBEDAAN ARAH ALIRAN UNTAI HEAT SINK SYSTEM. Kebutuhan energi listrik di Indonesia sudah saatnya didukung oleh energi nuklir, sesuai dengan rencana BATAN untuk membangun Reaktor Daya Eksperimental (RDE). Teknologi keselamatan pasif pada generasi reaktor lanjut sangat diprioritaskan untuk keselamatan operasi kondisi normal maupun tidak normal. Penelitian tentang sistem pasif telah dilakukan dengan eksperimen sirkulasi alam menggunakan untai Fasilitas Simulasi Sistem Pasif (untai FASSIP) di laboratorium Termohidrolika, Pusat Teknologi dan Keselamatan Reaktor Nuklir (PTKRN). Untai FASSIP terdiri dari untai rektangular dan untai Heat Sink System (HSS). Untai HSS berfungsi untuk menyerap kalor dari untai rektangular dan membuang ke lingkungan. Tujuan penelitian adalah untuk menganalisis kemampuan pertukaran kalor di tangki cooler berdasarkan perbedaan aliran untai HSS. Metodologi yang digunakan adalah melakukan eksperimen sirkulasi alam, membuat grafik, melakukan perhitungan jumlah kalor yang diserap HSS dan melakukan analisis kemampuan pertukaran kalor di tangki cooler berdasarkan perbedaan aliran HSS. Data hasil eksperimen dan perhitungan diperoleh analisis kemampuan pertukaran kalor di tangki cooler oleh untai HSS aliran berlawanan arah (counter flow) lebih besar dibandingkan dengan untai HSS aliran searah (parallel flow). Nilai jumlah kalor yang diserap di tangki cooler paling besar adalah 16,585 kW yaitu terjadi pada daya pemanas di tangki heater sebesar 4,24 kW dan laju aliran volumetrik rata-rata di untai HSS sebesar 42,70 LPM pada HSS aliran berlawanan arah. Kata Kunci : untai FASSIP, Heat Sink System, laju aliran volumetrik, perpindahan kalor

P-14 DEVELOPMENT OF INSTRUMENTATION AND CONTROL SYSTEM AT CREEP MACHINE USING LABVIEW SOFTWARE

Kussigit Santosa. Sudarno. Agus Nur Rahman, Kiswanta, Deswandri, Geni Rina Sunaryo Pusat Teknologi Dan Keselamatan Reaktor Nuklir - BATAN

Kawasan Puspiptek Gd 80. Serpong 15310. Tangerang Selatan Email: [email protected]

ABSTRACT DEVELOPMENT OF INSTRUMENTATION AND CONTROL SYSTEM AT CREEP MACHINE USING LABVIEW SOFTWARE. Creep machine on Mechanical Test Facility Laboratory (FUM) BPFKR is an important test tool that one of its functions is to predict the remaining life a component of the installation of nuclear power plants. This estimated life expectancy can be done by studying the characteristics of the elongation properties of the component material to the pressure or load on it at high temperatures (40% of the melting point). For operating temperature determination, this is still done manually, either during calibration or at the time of setting. The integration of this temperature control system in the computerized developed control system is needed to improve the system performance. The purpose of this integration is to simplify and improve the efficiency of data acquisition and time efficiency so that further research can be more easily done. The research methodology used consists of four stages. The first stage is study the existing creeps design and parameters. The second step is to determine the parameters which will be controlled to facilitate further processing. The third stage is determining the electronic module to be used in signal processing used. The modules are Ni 9074, Ni 9213 and Ni 9476. The last step is to create a data acquisition system by assembling the selected modules with the Lab VIEW software. The results of this activity is an easier control system, integrated data processing which has an error rate of 0.04%, then in the future processing of data and further control will be easier. Keywords: Creep machine, LabVIEW, Temperature control.

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P-15 SENSITIVITY AND UNCERTAINTY ANALYSIS OF J INTEGRAL CALCULATION OF THE RELIABILITY ASSESSMENT OF REACTOR PRESSURE VESSEL

Entin Hartini 1, Roziq Himawan 1, Abdul Hafid 1, Deswandri 1, Geni Rina Sunaryo1

1)Center for Nuclear Reactor Technology and Safety, Kawasan Puspiptek, Tangerang Selatan, 15310 Email: [email protected]

ABSTRACT SENSITIVITY AND UNCERTAINTY ANALYSIS OF J INTEGRAL CALCULATION OF THE RELIABILITY ASSESSMENT OF REACTOR PRESSURE VESSEL. The structural reliability of the reactor pressure vessel (RPV) is one aspect that must be considered in the safety analysis of the reactor. J integral calculation were conducted for structure reliability due to the cracks presence. In understanding the uncertainties that affect the output of the analysis, the evaluation of the uncertainty and sensitivity of the input variables need to be done. In the calculation of J integral input uncertainties variable involving physical variables for the loading condition that the internal pressure and material properties. The purpose of this study is to conduct a sensitivity analysis then to compare influence of the uncertainty of the outcome variable to output. RPV J integral calculation in 2D with initial crack modeled using MSC MARC. The calculation of input uncertainty was used simulation-based probabilistic density function (PDF). Then a sensitivity analysis using a variant of the conditional expectation was performed. The obtained results are stress intensity factor (SIF) to include the uncertainty of the load input will first reach the limit value of the fracture toughness (100 MPa m0.5) compared with the input uncertainty elastisity modulus. Based on the evaluation of the sensitivity value, the uncertainty of the load input heavily influence the integral J by 93.86% compared to the uncertainty of input on elasticity modulus. Keywords: Uncertainty, sensitivity, J integral, MSC MARC

P-16 MODIFIKASI PRE-HEATER MENJADI PRE-COOLER PADA UNTAI UJI BETA

Joko Prasetio W, Dedy H, Mulya Juarsa, Edy Sumarno, Deswandri, Geni R. Sunaryo

Pusat Teknologi Dan Keselamatan Reaktor Nuklir - BATAN Email: [email protected]

ABSTRAK Modifikasi Pre-Heater Menjadi Pre-Cooler Pada Untai Uji BETA. Fasilitas FASSIP-01 mempunyai komponen pre-heater dengan ukuran panjang 800 mm dengan diameter 10 inchi yang mempunyai kapasitas 50 kW dengan 10 batang pemanas preheater digunakan pada FASSIP-01 untuk memanaskan sistem aliran, dengan adanya perkembangan penelitian maka fasilitas BETA sudah tidak dioperasikan lagi, sedangkan dengan adanya fasilitas untai FASSIP-01 baru untai tersebut memerlukan sistem pendingin, dan untuk memanfaatkan pre-heater pada FASSIP-01 maka pre-heater dimodifikasi menjadi pre-cooler dengan sistem pendingin spiral yang dihubungkan dengan sistem Heat Sink System (HSS) pada untai FASSIP-01. Sehingga perlu dilakukan penelitian untuk mengetahui karakteristik pendingin secara pasif. Maka pada tahun 2015 dibuat Untai FASSIP-01untuk mempelajari karakteristik pendingin secara pasif. Kegiatan yang dilakukan pada tahun 2016 adalah memodifikasi pre-cooler dari sistem kontak tak langsung menjadi kontak langsung dan mengoperasikan Untai FASSIP-01dengan tujuan untuk memperoleh data pengukuran temperatur, daya listrik, frekuensi dan laju alir fluida yang diperlukan untuk mendukung penelitian sistem keselamatan pasif. Modifikasi pre-cooler dari kontak tak langsung menjadi kontak langsung telah berhasil menurunkan temperatur yang semula 24⁰C menjadi dibawah 10⁰C sesuai dengan yang diinginkan. Kata kunci: Modifikasi, pre-heater, pre cooler, sirkulasi alam, pendinginan

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P-17 RANCANG BANGUN ALAT PENGUKUR TEKANAN PADA UNTAI HSS FASSIP BERBASIS LABVIEW

Sumantri Hatmoko1, Kussigit santosa 1, Agus Nur Rachman 1, G. Bambang Heru1, Deswandri1, Geni Rina

Sunaryo1 1)Pusat Teknologi dan Keselamatan Reaktor Nuklir, BATAN, Kawasan Puspitek, Tangerang Selatan,15314

email: [email protected]

ABSTRAK RANCANG BANGUN ALAT PENGUKUR TEKANAN PADA UNTAI HSS FASSIP BERBASIS LABVIEW. Pada lab termohidrolika PTKRN BATAN telah dibangun Fasilitas Untai FASSIP. Fasilitas Untai FASSIP digunakan untuk eksperimental penelitian keselamatan PLTN berbasis pada sistem pasif. Untai FASSIP terdiri dari Untai Rectangular dan HSS. Pada Fasilitas Untai FASSIP ada beberapa parameter yang diukur yaitu laju alir dan temperatur. Alat ukur temperatur dan laju alir telah dibuat dan dipasang pada Fasilitas Untai Rectangular FASSIP. Pada penelitian selanjutnya parameter tekanan diperlukan dalam untai HSS FASSIP agar tekanan dapat diukur, diolah, disimpan dan ditampilkan dalam bentuk data realtime maka perlu dibuat rancang bangun alat pengukur tekanan pada untai HSS FASSIP berbasis labview. Untuk mengetahui perubahan tekanan pada untai HSS FASSIP maka digunakan pressure sensor U5100 measurement specialties yang dikoneksikan dengan modul National Instrument 9203 didalam port chassis DAQ NI 9188. Hasil penelitian ini adalah alat pengukur tekanan terpasang pada untai HSS FASSIP berbasis labview yang dapat menyimpan data hasil akuisisi secara komputerisasi dan realtime. Kata kunci: FASSIP, Labview, Tekanan

P-18 EFFECT OF CHLORIDE AND COPPER ION FOR REACTOR TANK MATERIAL INTEGRITY OF RSG GAS PRIMARY COOLING SYSTEM

Rahayu Kusumastuti1, Sumaryo2, Sriyono1, Sofia Loren1, M.Subekti1, Geni Rina Sunaryo1

1 Center for Nuclear Reactor Technology and Safety 2 Center for Science and Advanced Materials Technology

PUSPIPTEK Area 71 Building, Setu, Tangerang Selatan 15310 email: [email protected]

ABSTRACT

EFFECT OF CHLORIDE ION AND COPPER ION FOR TANK MATERIAL INTEGRITY OF RSG GAS PRIMARY COOLING SYSTEM. RSG GAS as Multipurpose Reactor designed to operate for 30 years, now RSG GAS has been operating for approximately 29 years. Al2Mg3 as a reactor tank material has a weakness in its use, that is not resistant for environment that contains of chloride ions and copper ion. The purpose of this research was to know the impact of Cl- and Cu2+for tank material integrity of the RSG GAS primary cooling system. The analysis is done by simulation on Al2Mg3 corrosion in Cl- and Cu2+ solution using Potensiogalvanostate IG & G. The results showed that the concentration of chloride ion in the primary cooling water that greater than 6 ppm have potentially to degrade the integrity of reactor tank material. The existence of copper ions in the primary cooling water will accelerate the damage to the reactor tank. Material integrity of RSG GAS reactor tank can be maintained by limiting the concentration of chloride and copper ions in each of the primary cooling water in accordance with the standard requirement. Keywords: integrity, the reactor tank material, the primary cooling water, RSG-GAS.

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P-19 MONITORING TEGANGAN LDR MENGGUNAKAN ARDUINO MEGA-2560 BERBASIS LabVIEW UNTUK PENGUKURAN KERAPATAN AEROSOL

G. Bambang Heru K, Alim Mardi, Joko P, Edy S, Deswandri, Geni R Sunaryo

Pusat Teknologi dan Keselamatan Reaktor Nuklir email: [email protected]

ABSTRAK

MONITORING TEGANGAN LDR MENGGUNAKAN ARDUINO MEGA-2560 BERBASIS LabVIEW UNTUK PENGUKURAN KERAPATAN AEROSOL. Pendinginan eksternal maupun internal sungkup reaktor dapat disimulasikan menggunakan FESPECo. DAS-NI sebagai sistem instrumentasi FESPECo dikembangkan guna memenuhi kebutuhan berbagai parameter pengukuran yang diperlukan, salah satunya adalah pengukuran kerapatan aerosol. Pada kegiatan ini dilakukan pembuatan program monitoring tegangan LDR yang digunakan sebagai sensor cahaya pada sistem pengukuran kerapatan aerosol. Tujuan kegiatan untuk memonitor tegangan LDR sebagai fungsi kerapatan aerosol yang dapat dipantau secara real time pada front panel LabVIEW. Dalam setiap tahapan simulasi, tegangan LDR dipantau berdasarkan intensitas cahaya yang diterima. Signal tegangan LDR tersebut dikondisikan dengan modul arduino mega-2560, selanjutnya dibaca menggunakan program aplikasi LabVIEW. Hasil eksekusi program menunjukkan tegangan LDR pada setiap tahapan simulasi dapat dipantau secara real time pada front panel LabVIEW. Dengan demikian program monitoring tegangan LDR dapat diaplikasikan pada sistem pengukuran kerapatan aerosol. Kata kunci: FESPECo, DAS-NI, LDR, aerosol, LabVIEW.

P-20 PEMBUATAN PROGRAM AKUISISI DATA PADA FASILITAS SIMULASI SISTEM PASIF (FASSIP)

Agus Nur Rachman1, Kussigit Santosa1, Sudarno1, Mulya Juarsa1, Deswandri1, Geni Rina S1

1 Pusat Teknologi dan Keselamatan Reaktor Nuklir, Gedung 80 Kawasan Puspiptek Serpong, Tangerang 15310 email: [email protected]

ABSTRAK

PEMBUATAN PROGRAM AKUISISI DATA PADA FASILITAS SIMULASI SISTEM PASIF (FASSIP). Kecelakaan reaktor nuklir Fukushima di Jepang terjadi akibat sistem pengambil panas darurat tidak dapat berfungsi. Generator yang digunakan untuk menjalankan pompa air pendingin terendam air akibat Tsunami. Hal ini menunjukkan pentingnya sistem pendingin dengan sistem pasif untuk diterapkan dalam reaktor. FASSIP dibangun untuk mempelajari fenomena sirkulasi alamiah (Natural Circulation) pada sistem pendingin reaktor. Fasilitas ini dilengkapi dengan sensor suhu, tekanan dan aliran untuk mengamati perubahan parameter yang terjadi di dalam untai FASSIP. Untuk mengamati perubahan fisika yang terjadi pada untai FASSIP digunakan sensor termokopel, pressure trasnduser, dan flowmeter. Seluruh sensor ini dihubungkan pada satu sistem instrumentasi menggunakan perangkat keras dari National Instrument. Untuk menampilkan hasil pengukuran dibuat sistem antarmuka menggunakan perangkat lunak LabVIEW. Uji fungsi dilapangan menunjukkan sistem instrumentasi yang telah dibangun mampu mengukur dan mengakuisisi fenomena fisika pada untai FASSIP. Sistem antarmuka yang telah dibuat mampu menampilkan hasil pengukuran baik dalam bentuk numerik, tabel dan grafik. Berdasarkan hasil uji fungsi sistem instrumentasi dan antarmuka pada untai uji FASSIP mampu mengakusisi data dari 71 termokopel, 2 Flowmeter, dan 2 pressure transduser secara real time, serta menyimpan data tersebut pada komputer untuk melakukan analisa terhadap hasil eksperimen. Kata kunci: Sistem Pasif, Akuisisi Data, Antarmuka, LabVIEW

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P-21 ANALISIS TERMOHIDROLIK TERAS REAKTOR TRIGA 2000 BANDUNG BERELEMEN BAKAR TIPE PELAT MENGGUNAKAN PROGRAM CFD

Reinaldy Nazar, Sudjatmi KA, K. Kamajaya

Pusat Sains dan Teknologi Nuklir Terapan - BATAN Jln. Tamansari No. 71, Bandung, 40123


ABSTRAK ANALISIS TERMOHIDROLIK TERAS REAKTOR TRIGA 2000 BANDUNG BERELEMEN BAKAR TIPE PELAT MENGGUNAKAN PROGRAM CFD. Mengingat tidak diproduksinya lagi elemen bakar TRIGA oleh produsen elemen bakar TRIGA General Atomic perlu diusahakan suatu solusi agar reaktor TRIGA 2000 dapat tetap beroperasi. Salah satu solusi yang dapat ditawarkan terkait permasalahan di atas adalah dengan melakukan penggantian tipe elemen bakar. Pada studi ini telah dianalisis penggunaan elemen bakar tipe pelat sejenis yang terpasang di RSG-GAS, untuk dipasangkan pada teras reaktor TRIGA 2000 Bandung. Berdasarkan hasil penelitian yang telah dilakukan, dengan menggunakan program komputer CFD, diketahui bahwa reaktor TRIGA berelemen bakar tipe pelat tidak dapat dioperasikan pada daya 2000 kW dengan menggunakan moda pendinginan konveksi alamiah seperti yang ada saat ini. Untuk itu dilakukan dengan moda pendinginan konveksi paksa. Hasil analisis konveksi paksa menunjukkan dengan menggunakan laju alir pendingin pompa 50 kg/s dan temperaturnya bervariasi 35 oC, 35,5 oC dan 36 oC, diperoleh temperatur permukaan pelat elemen bakar antara 110,37 oC – 111,27 oC dan temperatur pendinginnya pada posisi terkait antara 61,03 oC – 61,95 oC. Temperatur permukaan pelat elemen bakar ini mendekati temperatur saturasi dan tentunya telah mulai terjadi pendidihan inti, sehingga penggunaan laju alir pendingin masuk teras 50 kg/s perlu dihindari. Temperatur permukaan pelat elemen bakar mulai menjauhi temperatur saturasi jika digunakan laju alir pendingin lebih besar dari 65 kg/s, dimana diperoleh temperatur permukaan pelat elemen bakar 96,65 oC dan temperatur pendinginnya pada posisi terkait 54,38 oC. Kata kunci: reaktor TRIGA 2000, elemen bakar tipe pelat, termohidrolik, program CFD

P-22 EFFECT OF TEMPERATURE TO ADSORPTION CAPACITY AND DISTRIBUTION COEFFICIENT ON RARE EARTH ELEMENTS ADSORPTION (Y, Dy, Gd) USING SIR

Dwi Biyantoro1, Agus Taftazani1, Aswati Mindaryani2, Supranto2, Nofriady Aziz2 Pusat Sains dan Teknologi Akselarator – BATAN1 , Universitas Gadjah Mada2

Jl. Babarsari PO BOX 6101 YKBB Yogyakarta Telp. (0274)48085, 489816 ; Fax (0274) 489715


ABSTRACT EFFECT OF TEMPERATURE TO ADSORPTION CAPACITY AND DISTRIBUTION COEFFICIENT ON RARE EARTH ELEMENTS ADSORPTION (Y, Dy, Gd) USING SIR. The use of Rare Earth Elements (REEs) like element of yttrium (Y) as a superconducting material requires a purity of more than 90% so it needs to increase the purity of Y from the separation and purification process. The purpose of this research is to study the separation process of REEs (Y, Gd, Dy) elements from RE Hydroxide (RE(OH)3) using Solvent Impregnated Resins (SIR). This research was conducted on variation of SIR composition, temperature variation of adsorption process, determination of equilibrium equation and kinetic sorption occurring in SIR adsorption. Based on the calculation result, the most effective SIR composition for REEs separation is 0.75 g, the equilibrium equation for Y, Gd and Dy approximates the Henry equilibrium model and the pseudo kinetic model of the reaction order Y, Gd, and Dy is approximated by the pseudo reaction of second order. The result of separation of RE(OH)3 with SIR is said to be effective from other method because purity is obtained that is 96.73% and qualify as a superconductor material. Keywords: SIR, Amberlite XAD-16, REEs, TBP, D2EHPA

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P-23 ANALISIS DISTRIBUSI TEMPERATUR KANAL TERPANAS TERAS REAKTOR TRIGA BANDUNG BERBAHAN BAKAR PELAT DENGAN PROGRAM FLUENT

V.Indriati Sri Wardhani1, Henky P. Rahardjo2 dan Surip Widodo3

1,2Pusat Sains dan Teknologi Nuklir Terapan- Batan- Bandung 3Pusat Teknologi Reaktor dan Keselamatan Nuklir-kawasan Puspiptek- Serpong


ABSTRAK ANALISIS DISTRIBUSI TEMPERATUR KANAL TERPANAS TERAS REAKTOR TRIGA BANDUNG BERBAHAN BAKAR PELAT DENGAN PROGRAM FLUENT. Unjuk kerja sistem pendinginan dalam teras reaktor merupakan salah satu faktor yang harus dipertimbangkan untuk memberikan jaminan operasi reaktor secara aman. Proses perpindahan panas dan berbagai parameter-parameter aliran yang terjadi di teras reaktor seperti distribusi temperatur, distribusi tekanan, distribusi kecepatan, laju alir massa dan pola aliran fluida yang terjadi perlu dilakukan analisis. Dalam rangka mendukung program konversi reaktor TRIGA dari bahan bakar silinder menjadi bahan bakar pelat akan dilakukan analisis dari berbagai parameter di atas untuk mendukung unjuk kerja sistem pendinginannya. Analisis dilakukan menggunakan program FLUENT dengan membuat suatu pemodelan satu bundle bahan bakar yang terpanas. Program ini membantu untuk memecahkan persamaan matematis yang merumuskan proses dinamika fluida dalam menggambarkan fenomena aliran fluida yang terjadi. Dalam makalah ini dilakukan analisis konservatif pada bundle bahan bakar terpanas dalam teras reaktor meliputi distribusi temperatur, distribusi tekanan dan laju alir fluida pendingin yang melewati kanal atau celah di antara pelat dalam satu bundel bahan bakar. Distribusi temperature dan distribusi tekanan yang diperoleh dari analisis tersebut dapat dipergunakan untuk memprediksi tingkat keadaan termohidrolik pendingin teras reaktor TRIGA pelat pada saat satu fasa atau dua fasa, yang sangat berarti bagi unjuk kerja pendingin. Diharapkan hasilnya dapat mewakili unjuk kerja teras reaktor TRIGA berbahan bakar pelat secara keseluruhan sehingga dapat dinyatakan bahwa operasi reaktor berjalan dengan aman. Kata kunci : konversi, pelat, bundle, bahan bakar, pendingin, kanal, konservatif.

P-24 STUDI PERENCANAAN ENERGI KELISTRIKAN KALIMANTAN TIMUR DENGAN OPSI PLTN JENIS SMR

Wiku Lulus Widodo1, Rizki Firmansyah Setya Budi1 PKSEN-BATAN, Kuningan Barat, Jakarta Selatan 12710


ABSTRAK STUDI PERENCANAAN ENERGI KELISTRIKAN KALIMANTAN TIMUR DENGAN OPSI NUKLIR. Telah dilakukan studi perencanaan energi kelistrikan untuk wilayah Kalimantan Timur hingga tahun 2050. Batubara sebagai bahan bakar untuk menghasilkan energi listrik dapat digunakan untuk memenuhi kebutuhan listrik di Kalimantan Timur. Kondisi sumber daya energi batubara yang melimpah di Kalimantan Timur belum banyak dimanfaatkan untuk sektor kelistrikan. Studi ini bertujuan untuk menganalisis dan memproyeksikan kondisi kelistrikan di Kalimantan Timur hingga tahun 2050 dengan opsi nuklir. Metode yang digunakan dalam studi ini adalah mengumpulkan data kelistrikan Kalimantan Timur, mendesain rantai energi untuk MESSAGE dan menghitung energi listrik yang diproduksi setiap pembangkit listrik hingga tahun 2050. PLTN yang digunakan adalah SMR (Small Medium Reactor) dengan kapasitas 200 MW. Hasil perhitungan diperoleh dengan kebutuhan listrik Kalimantan Timur pada tahun 2020 diperkirakan sebesar 4,8 Twh dan naik pada tahun 2050 hingga mencapai 28,5 TWh. Hasil skenario non nuklir menunjukkan bahwa PLTU diproyeksikan akan mendominasi hingga pada tahun 2050 mencapai 18,41 Twh. Hasil skenario nuklir menunjukkan bahwa pada tahun 2040 PLTN SMR akan berperan untuk memenuhi kebutuhan beban dasar maka dengan produksi listrik sebesar 0,88 Twh atau pangsa sebesar 4,7%. Adapun untuk tahun 2050 maka PLTN SMR akan memproduksi 2,63 Twh atau pangsa sebesar 9,3% dari total listrik yang diproduksi.PLTN SMR akan berfungsi mengurangi pangsa PLTU batubara sehingga bermanfaat untuk menekan pertumbuhan emisi karbon. Kata kunci : MESSAGE, SMR , Kalimantan Timur, Metodologi INPRO, Energi

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P-25 ANALISIS KEKUATAN MEKANIK ALAT BANTU ULTRASONIK UNTUK PEMERIKSAAN BEAM TUBE RSG-GAS

Dedy Haryanto*, Almira Citra Amelia*, Aep Saepudin Catur **

*Pusat Teknologi dan Keselamatan Reaktor Nuklir **Pusat Reaktor Serba Guna

ABSTRAK

ANALISIS KEKUATAN MEKANIK ALAT BANTU ULTRASONIK UNTUK PEMERIKSAAN BEAM TUBE RSG-GAS. Reaktor Serba Guna G.A Siwabessy (RSG-GAS) memiliki fasilitas beam tube yang berfungsi sebagai tabung berkas tempat terjadinya proses iradiasi. Salah satu langkah perawatan terhadap beam tube yang akan dilakukan adalah pemeriksaan kondisi kelayakan beam tube menggunakan metode ultrasonik. Oleh karena itu maka perlu didesain alat bantu ultrasonik untuk mendukung pelaksanaan pemeriksaan pada beam tube. Alat bantu tersebut didesain mampu membawa probe ultrasonik dan meletakkan probe tepat pada jarak yang diijinkan sesuai dengan ketentuan pemeriksaan dengan metode ultrasonik. Analisis kekuatan mekanik alat bantu harus dilakukan setelah desain alat diperoleh dan sebelum dipabrikasi. Analisis ini bertujuan untuk mengetahui kelayakan dan keamanan alat bantu ketika digunakan untuk pemeriksaan beam tube. Alat bantu didesain menggunakan material Aluminium Alloy 1050 (Al 1050) dan analisis dengan metode pengujian secara simulasi menggunakan software CATIA. Tahapan yang dilakukan dalam menganalisis kekuatan mekanik yaitu ; pembuatan model 3-dimensi yang dilengkapi dengan sifat mekanik material yang digunakan yaitu Al 1050, memberikan restraint dan beban pada model 3-dimensi dan melakukan pengujian secara simulasi. Hasil pengujian pada perenggangan (translational displacement) 2,69 mm menimbulkan tegangan mekanik (von mises stress) sebesar 36,5 Mpa dengan gaya yang diberikan sebesar 125 N. Hasil tersebut dinilai baik karena pada perenggangan maksimal yang akan terjadi pada saat alat bantu digunakan tegangan mekanik yang terjadi jauh lebih kecil daripada yield strength material yang digunakan dan masih berada pada area elastisnya. Sehingga desain alat bantu ultrasonik dapat dipabrikasi dan digunakan untuk mendukung pemeriksaan beam tube dengan metode ultrasonik. Kata kunci :Beam tube, alat bantu ultrasonik, translational displacement, tegangan mekanik.

P-26 PEMBUATAN SILIKON KARBIDA MONOLITIK DAN KARAKTERISTIKNYA

Futichah, Deni Mustika, Heri Hardiyanti, Pranjono, Isfandi, Jan Setiawan PTBBN-BATAN, Kawasan PUSPIPTEK Serpong

ABSTRAK

PEMBUATAN SILIKON KARBIDA MONOLITIK DAN KARAKTERISTIKNYA. Pembuatan SiC monolitik sebagai bahan pelapis pertama pada kelongsong bahan bakar nuklir berbasis komposit SiC telah dilakukan. Pada temperatur tinggi, SiC monolitik merupakan keramik yang mempunyai unjuk kerja yang lebih tinggi bila dibandingkan dengan logam. Dalam penelitian ini dilakukan karakterisasi mekanik dan morfologi pada SiC monolitik yang dibuat melalui proses metalurgi serbuk dengan tekanan dan waktu kompaksi yang berbeda dan dengan proses sinter pada temperatur tertentu. Dari hasil karakterisasi diharapkan diperoleh data parameter pembuatan SiC monolitik yang terbaik. Bahan yang digunakan dalam penelitian ini adalah serbuk SiC dan larutan pengikat yang terdiri dari larutan toluen dan polycarbosilane dengan komposisi yang bervariasi. Serbuk SiC ditimbang kemudian dicampur dengan larutan binder, selanjutnya dikompakan menjadi pelet. Pelet hasil kompaksi disinter dalam tungku sinter mini pada temperatur 1300 oC selama 3 jam. Pelet SiC monolitik hasil sinter dikarakterisasi untuk mengetahui densitasnya menggunakan alat autopicnometer, struktur mikro SiC monolitik menggunakan mikroskop optik dan uji kekerasan mikro untuk mengetahui kekerasannya dengan microhardness tester. Semua uji yang dilakukan bertujuan untuk memperoleh kondisi parameter proses penahanan tekanan kompaksi yang optimum. Hasil sinter SiC monolitik yang telah dibuat mempunyai struktur mikro dimana SiC telah berdifusi. Hasil yang diperoleh menunjukkan bahwa proses kompaksi serbuk SiC relatif baik pada tekanan 10 kN dengan densitas SiC monolitik sebesar (3,4± 0,1) g/cm3 dan nilai kekerasan rata-rata tertinggi sebesar 17,4 HVN. Karakteristik SiC monolitik yang diperoleh lebih rendah dibandingkan nilai secara teoritiknya dan sebagai lapisan pertama pada kelongsong komposit SiC belum dapat disimpulkan kinerjanya sebagai kelongsong bahan bakar nuklir. Kata kunci: SiC monolitik, keramik, komposit, polycarbosilane, zirkaloi.

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P-27 PROYEKSI NERACA ENERGI INDONESIA HINGGA TAHUN 2060

Edwaren Liun Pusat Kajian Sistem Energi Nuklir, BATAN, Jl Kuningan Barat, Mp. Prapatan, Jakarta 12710


ABSTRAK PROYEKSI NERACA ENERGI INDONESIA HINGGA TAHUN 2060. Indonesia masih berada dalam pertumbuhan permintaan energi yang tinggi yang dipacu oleh pertumbuhan penduduk, pertumbuhan ekonomi perubahan gaya hidup dan urbanisasi. Hingga tahun 2060 pertumbuhan masih berlanjut dimana penduduk Indonesia telah mencapai 372 juta jiwa. Tulisan ini bertujuan untuk menciptakan wawasan terhadap pilihan teknologi energi untuk mencapai sistem pembangkit listrik yang ekonomis. Metode yang digunakan adalah pengolahan data sekunder berupa statistik makro-ekonomi dan energi, proyeksi permintaan energi, dan analisis kecenderungan kurva data statistik terkait permintaan energi. Data disusun dan diintegrasikan di dalam model sesuai format yang tersedia dan pengolahan dengan program pemodelan energi MASSAGE. Hasil yang diperoleh adalah proyeksi permintaaan energi dan strategi penyediaan hingga tahun 2060 terutama energi listrik. Hasil kuantitatif yang disajikan dalam makalah ini terdiri dari peringkat alternatif teknologi listrik dan ranking portofolio listrik. Hasil yang diperoleh adalah proyeksi jangka panjang hingga tahun 2060 dimana pada tahun tersebut kebutuhan total energi Indonesia akan mencapai 7,12 milyar BOE. Sedangkan suplai energi listrik yang dibutuhkan akan mencapai 296 ribu MWy, terdiri dari 235 ribu MWy 61 ribu MWy masing-masing untuk wilayah Jawa-Bali, Madura dan Sumatera, dan wilayah Indonesia lainnya. Secara keseluruhan energi paling dominan yang akan muncul untuk pembangkit listrik adalah dari jenis bahan bakar fosil terutama gas dan batubara. Kata kunci: Jumlah penduduk, permintaan energi, proyeksi jangka panjang, optimasi, sistem kelistrikan.

P-28 MANAJEMEN KONSTRUKSI REAKTOR DAYA EKPERIMENTAL

Mudjiono1, Erlan Dewita1, Dedy Priambodo1

1 Pusat Kajian Sistem Energi Nuklir, Jl. Kuningan Barat, Mampang Prapatan, Jakarta 12710 email: [email protected]

ABSTRAK MANAJEMEN KONSTRUKSI REAKTOR DAYA EKSPERIMENTAL. Proyek konstruksi Pembangkit Listrik Tenaga Nuklir (PLTN) akan menghadapi berbagai tantangan baik yang bersifat teknis maupun non teknis, untuk itu dibutuhkan manajemen pengelolaan proyek yang baik dan benar untuk menghadapi berbagai tantangan tersebut. Telah dilakukan kajian tentang manajemen konstruksi Reaktor Daya Eksperimental (RDE). Kajian ini bertujuan untuk memperoleh sistem pengaturan dan pengelolaan dalam manajemen konstruksi RDE sedemikian rupa hingga diperoleh hasil yang optimal sesuai dengan rencana dan spesifikasi yang ditetapkan. Metodologi yang digunakan adalah studi pustaka, pengumpulan data sekunder dan pelaksanaan FGD (focus group discussion) dengan menghadirkan pakar dibidangnya. Hasil kajian diperoleh bahwa Keberhasilan pelaksanaan manajemen konstruksi RDE memerlukan kesinambungan kegiatan pada tahap perencanaan, pelaksanaan konstruksi dan pengawasan terhadap pekerjaan. Disisi lain diperlukan dokumen perencanaan yang lengkap, pengawasan yang ketat dan pengendalian proyek secara terstruktur serta didukung dengan sumber daya manusia yang handal sehingga kegiatan proyek dilakukan secara profesional. Kata kunci: manajemen, konstruksi, RDE

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P-29 EFFECT OF HEAT TREATMENT ON THE STRENGTH OF AUSTENITIC STAINLESS STEEL SS304

Andryansyah, Mudi H, Arismunandar P S, Darlis, Dwijo M, Deswandri, Geni RS Center for Nuclear Reactor Technology and Safety - National Nuclear Energy Agency of Indonesia, Puspiptek

Gd.80, Tangerang Selatan email: [email protected]

ABSTRACT

EFFECT OF HEAT TREATMENT ON THE STRENGTH OF AUSTENITIC STAINLESS STEEL SS304. Stainless steel is used in the nuclear industry and nuclear power plants because it has good corrosion resistance. In its usage, occasional heating process results a sensitization. Past research shows that it is possible to restore the microstructure of austenitic stainless steels that have undergone a process of sensitization to the initial conditions. To see the possibility of it then do some heat treatment processes on stainless steel SS304 in the hope it can also improve its mechanical properties. The samples are heated at a temperature varies from 500OC to 1100OC. The heating time varies and so does the cooling process. The repair process is done at a temperature of 1100OC in the furnace with cooling variation. The process of heating also made with LPG flame to the melting point and followed by cooling using ice water. Metal strength represented by the hardness testing that have a direct correlation to the tensile strength of the metals. From the research that has been done, it can be concluded that it is not possible to upgrade the hardness of SS304 with heat treatment process. Each heating process followed by varies cooling of SS304 produce the hardness in the same or decreased. The results can also be used to conclude that the welding process on the SS304 will always result in a decrease in the strength of the metal parts that are affected by the welding heat. Keywords: heat treatment, improved strength, SS304.

P-30 PENENTUAN PENUMBRA PADA RADIOGRAFI BENDA BERGERAK

Zaenal Abidin, Angga Fernando, Djoko Marjanto Jurusan Teknofisika Nuklir, Sekolah Tinggi Teknologi Nuklir – BATAN

Jl. Babarsari PO BOX 6101 / YKBB Yogyakarta 55281, Telp. (0274)484085; Fax : (0274)489715 Email: [email protected]

ABSTRAK

Penentuan Penumbra Pada Radiografi Benda Bergerak. Pergeseran benda uji adalah satu dari faktor yang mempengaruhi kualitas film radiografi, standart ASME V mandatory appendix I menjelaskan tentang persyaratan yang harus dipenuhi dalam pengendalian pergeseran radiografi. Penelitian mengenai benda bergerak dilakukan untuk mengetahui pengaruh pergerakan terhadap kualitas film radiografi dalam bidang industri. Penelitian ini dilakukan dengan membuat benda uji yang dirancang untuk dapat bergerak secara horizontal. Pergerakan dilakukan dengan pergeseran sebesar 0,5 cm, 1 cm, 2 cm, 3 cm, dan 4 cm dengan variasi tegangan 120 kV, 140 kV, 160 kV dan 180 kV. Tujuan dari penelitian ini adalah untuk menentukan ketidaktajaman geometri terhadap pengaruh pergeseran benda uji sesuai standart ASME V. Hasil penelitian tanpa pergeseran radiografi dengan memvariasi kV dapat mempengaruhi densitas dan kontras subyek pada film radiografi, sedangkan pengujian pergeseran radiografi, dengan memvariasi kV dan jarak pergeseran dapat mempengaruhi kontras subyek dan ketidaktajaman geometri. Pada percobaan pergeseran radiografi masih diterima dengan jarak pergeseran 0,5 cm - 3 cm pada tegangan 140 kV, 160 kV dan 180 kV dengan nilai Ug < 0,51 mm sesuai standart ASME. Kata Kunci : Pergeseran radiografi, ketidaktajaman pergeseran, cacat, ASME V, Radiografi.

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P-31 ASPEK DEMOGRAFI MENDUKUNG KEGIATAN PRA-SURVEI TAPAK PLTN DI BARELANG (BATAM, REMPANG, GALANG), KEPRI

June Mellawati1, Siti Alimah2 1Pusat Teknologi Keselamatan dan Metrologi Radiasi, BATAN,

Jl. Lebak Bulus Raya 49, Ps. Jumat PO Box 7043, Jaksel 12070 2Pusat Kajian Sistem Energi Nuklir, BATAN,

Jl. Kuningan Barat, Mampang Prapatan, Jaksel 12710, E-mail: [email protected]

ABSTRAK

ASPEK DEMOGRAFI MENDUKUNG KEGIATAN PRA-SURVEI TAPAK PLTN DI BARELANG (BATAM, REMPANG, GALANG), KEPRI. Aspek demografi (kependudukan) merupakan salah satu aspek yang perlu dipertimbangkan dalam kegiatan pra-survei tapak PLTN, karena dalam azas keselamatan teknologi nuklir harus mencakup keselamatan masyarakat dan lingkungannya. Dalam rangka menindaklanjuti SK kerjasama antara Batan dan BIFZA Nomor B 1741/KA/Batam/KS 0001/02/2015 dan Nomor 79/SPJ/KA/2/2015 telah dilakukan penelitian aspek kependudukan untuk mendukung kegiatan pra-survei tapak PLTN di wilayah Barelang Kepulauan Riau. Tujuan penelitian adalah memperoleh data base kependudukan di Barelang sebagai masukan para pemangku kepentingan (BIFZA) di Provinsi Kepulauan Riau terkait dengan rencana alokasi calon tapak PLTN di Batam. Metodologi meliputi pengumpulan data sekunder penduduk dari BPS Pusat dan daerah, konfirmasi lapangan untuk memperoleh data primer, evaluasi dan analisis data. Hasil penelitian menunjukkan bahwa jumlah penduduk di Pulau Rempang 1992 jiwa, Pulau Galang 1417 jiwa, Galang Baru 884 jiwa dan Pulau Batam 413.428 jiwa. Jumlah penduduk perempuan di Barelang (Kota Batam) pada 1999-2007 lebih banyak dibandingkan laki-laki dengan rasio 0,85 - 0,96, namun tahun 2008 - 2014 jumlah penduduk laki-laki lebih banyak dibandingkan perempuan dengan rasio 1,02 - 1,09. Kepadatan penduduk di Pulau Batam 996,21 jiwa/Km2, Pulau Rempang 12,07 jiwa/Km2, Pulau Galang 17,71 jiwa/Km2 dan Galang Baru 27,63 jiwa/Km2). Berdasarkan Perka BPS No. 37 Tahun 2010, Pulau Rempang, Galang dan Galang Baru mempunyai nilai/skor kepadatan 1 yang berarti kepadatannya sangat rendah. Berdasarkan data ini, Pulau Rempang, Galang dan Galang Baru berpeluang sebagai calon tapak dibandingkan Pulau Batam. Kata kunci: penduduk, tapak PLTN, Barelang

P-32 PERKIRAAN BIAYA EKSTERNAL DARI FASILITAS NUKLIR RDE MENGGUNAKAN SOFTWARE SIMPACT

Sufiana Solihat1, Wiku Lulus Widodo1 1) Pusat Kajian Sistem Energi Nuklir (PKSEN) – BATAN

Jalan Kuningan Barat, Mampang Prapatan, Jakarta Selatan 12710 email: [email protected]

ABSTRAK PERKIRAAN BIAYA EKSTERNAL DARI FASILITAS NUKLIR RDE. Telah dilakukan studi perhitungan biaya eksternal dari fasilitas nuklir RDE. Biaya eksternal yang dihitung adalah biaya kompensasi dari dampak kesehatan untuk area atau wilayah di sekitar fasilitas nuklir RDE. Dampak kesehatan difokuskan pada risiko kanker dan penyakit turunan spesifik yang diterima oleh masyarakat di sekitar fasilitas RDE akibat inhalasi radionuklida. Metode yang digunakan adalah membuat model instalasi RDE dengan menggunakan perangkat lunak SIMPACT. Hasil perhitungan SIMPACT berupa tingkat penyebaran radionuklida dan kemudian dikonversi menjadi harga kompensasi risiko kanker akibat penyebaran radionuklida tersebut. Hasil perhitungan menunjukkan bahwa biaya eksternal untuk risiko kanker fatal sebesar 0,00000185 US$/tahun, untuk risiko kanker non-fatal sebesar 0,00000482 US$/tahun, dan untuk risiko penyakit turunan spesifik sebesar 0,00001209 US$/tahun. Kata kunci: RDE, Biaya Eksternal, Kanker, Inhalasi, SIMPACT

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P-33 GAMBARAN PENERAPAN PENILAIAN DIRI DALAM PENCAPAIAN BUDAYA KESELAMATAN (STUDI KASUS DI BATAN)

Farida Tusafariah1, Deswandri2, Arie Budianti3

1 Pusat Teknologi Keselamatan dan Metrologi Radiasi-BATAN,Kawasan Nuklir Pasar Jumat Jl. Lebak Bulus Raya No.49, Jakarta 12440

2 Pusat Teknologi Keselamatan Reaktor Nuklir-BATAN, Kawasan Nuklir Serpong,Tangerang Selatan 3 Pusat Teknologi Limbah Radioaktif-BATAN, Kawasan Nuklir Serpong,Tangerang Selatan


ABSTRAK GAMBARANPENERAPAN PENILAIAN DIRI DALAM PENCAPAIAN BUDAYA KESELAMATAN (STUDY KASUS DI BATAN). BATAN sebagai fasilitas yang memanfaatkan tenaga nuklir dalam kegiatannya mempunyai risiko terjadinya kecelakaan yang disebabkan kondisi sikap dan perilaku baik individu maupun organisasi. Oleh sebab itu dalampengoperasian fasilitas nuklir harus menerapkan budaya keselamatan untuk meningkatkan kesadaran setiap individu akan pentingnya aspek keselamatan.Makalah ini bertujuan untuk menganalis penerapan budaya keselamatan unit kerja di BATAN, melalui penilaian diri berdasarkan Peraturan Kepala BATAN nomor 200/KA/X/2012 yang telah dilaksanakan dari tahun 2013-2016. Penilaian diri dilakukan menggunakan metode semi kuantitatif, dengan teknik pengumpulan data melalui kuesioner dan observasi. Analisis data dilakukan dengan pengumpulan data, reduksi data, dan penyajian data.Hasil penerapan penilaian diri budaya keselamatan di BATAN diperoleh nilai antara549-799 selama tahun 2013-2016. Hasil menggambarkan terjadinya peningkatan dari tahun 2013, 2014-2015, namun ada beberapa unit kerja yang mengalami penurunan di tahun 2016. Tetapi kepedulian terhadap keselamatan di unit kerja sudah meningkat, dan penilaian diri sudah dilakukan oleh semua unit kerja di BATAN. Program maupun kegiatan yang dilakukan dapat memperkuat budaya keselamatan, dan harus ditingkatkan lagi secara berkelanjutan.Target penilaian diri peringkat B secara umum sudah tercapai, sesuai yang diharapkan oleh Kepala BATAN dan komitmen bersama. Kata kunci: Penilaian diri, Metode kuesioner, Nilai peringkat, Potret Batan

P-34 ANALISIS POTENSI LIKUIFAKSI DI TAPAK REAKTOR DAYA EKSPERIMENTAL SERPONG Eko Rudi Iswanto1), Heri Syaeful2), Sriyana3)

1)3) PKSEN-BATAN, Jl. Kuningan Barat Mampang Prapatan, Jakarta Selatan 12710 2) PTBGN-BATAN, Jl. Lebak Bulus Raya Lebak Bulus, Jakarta Selatan 12440


ABSTRAK ANALISIS POTENSI LIKUIFAKSI DI TAPAK REAKTOR DAYA EKSPERIMENTAL SERPONG. Fenomena berubahnya sifat sedimen dari keadaan padat menjadi keadan cair yang disebabkan oleh tegangan geser pada saat gempa yang terjadi bolak balik disebut dengan likuifaksi. Likuifaksi dapat menyebabkan kerusakan berat hingga kegagalan struktur. Analisis potensi likuifaksi ini bertujuan untuk mengetahui potensi terjadinya likuifaksi pada lokasi RDE bangunan reaktor. Analisis dilakukan dengan metode simplified dengan menggunakan data Standard Penetration Test (SPT). Dari data tersebut, kemudian dapat dihitung nilai Cyclic Stress Ratio (CSR), nilai Cyclic Resistant Ratio (CRR) dan Safety Factor (SF). Berdasarkan analisis perhitungan yang dilakukan, disimpulkan bahwa area bangunan reaktor memiliki lapisan tanah yang berpotensi terlikuifaksi pada kedalaman 2 hingga 6 meter dan pada kedalaman 24 meter. Keyword: likuifaksi, Standard Penetration Test, Cyclic Stress Ratio, Cyclic Resistant Ratio

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P-35 IDENTIFIKASI KETERDAPATAN THORIUM PADA ENDAPAN LATERIT BAUKSIT DI DAERAH NANGA TAYAP – SANDAI, KABUPATEN KETAPANG,

KALIMANTAN BARAT

Widodo1, Putri Rahmawati1, Ngadenin1 1Pusat Teknologi Bahan Galian Nuklir, BATAN Jl. Lebak Bulus raya No 9 Pasar Jumat Jakarta Selatan


ABSTRAK Hasil survei uranium dan thorium regional di Kabupaten Ketapang, Propinsi Kalimantan Barat yang dilakukan menggunakan surveymeter gamma RS 125 mendapatkan zona anomali uranium dan thorium pada endapan laterit bauksit yang menempati wilayah batuan gunungapi Kerabai, granit Sukadana dan basalt Bunga di daerah Nanga Tayap dan Sandai. Wilayah Nanga Tayap dan Sandai merupakan daerah pertambangan laterit bauksit. Tujuan penelitian ini adalah mengetahui keterdapatan thorium pada endapan laterit bauksit karena penyebaran laterit bauksit di Indonesia cukup melimpah. Metode yang digunakan adalah dengan cara pembuatan sumur uji pada wilayah batuan gunungapi Kerabai, granit Sukadana dan basal Bunga, pengambilan sampel soil/batuan pada sumur uji dan analisis kadar unsur thorium dan aluminium. Hasil penelitian menyimpulkan bahwa kadar thorium tertinggi terdapat pada sumur uji di wilayah batuan basalt Bunga yaitu mencapai 115 ppm sedangkan yang terendah terdapat pada wilayah batuan gunungapi Kerabai yaitu 28 ppm. Di wilayah batuan granit Sukadana dan batuan basalt Bunga unsur thorium dan aluminium tidak berkorelasi tetapi di wilayah batuan gunungapi Kerabai berkorelasi cukup baik. Di wilayah granit Sukadana pengayaan thorium terjadi pada zona antara limonit dan saprock, sedangkan di wilayah batuan basalt Bunga terjadi pada zona saprolith dan di wilayah batuan gunungapi Kerabai terjadi pada zona limonit. Kata kunci: Thorium, aluminium, bauksit, Ketapang, Kalimantan Barat.

P-36 KAJIAN KESELAMATAN TAPAK RDE BERDASARKAN SURVEI PEDOLOGI DI KAWASAN PUSPIPTEK SERPONG, PROVINSI BANTEN

Hadi Suntoko, Heni Susiati

Pusat Kajian Sistem Energi Nuklir-BATAN,Jl Kuningan Barat, Mampang Prapatan, Jakarta, 12710 email: [email protected]

ABSTRAK KAJIAN KESELAMATAN TAPAK RDE BERDASARKAN SURVEI PEDOLOGI DI KAWASAN PUSPIPTEK SERPONG, PROVINSI BANTEN. Serpong telah memenuhi syarat menjadi tapak RDE sesuai Perka BAPETEN No 8/2013 tentang aspek kegempaan, yang salah satunya dilakukan kajian berdasarkan survei pedologi hingga radius 25 km, meliputi beberapa Kabupaten di Provinsi Banten. Pendataan pedologi memberikan informasi kondisi geologi permukaan melalui deskripsi tanah dan tata guna lahan yang menunjukkan jejak tektonik/ patahan yang dapat dikenali dengan perubahan posisi tanah. Tujuan survei adalah untuk kajian keselamatan tapak RDE dari karakteristik patahan terutama jejak patahan yang dapat dikenali dari kondisi geologi tanah/ pedologi. Hampir semua lokasi di permukaan bumi telah mengalami pelapukan batuan menjadi tanah akibat sifat fisik kimia, namun akan memberikan arti berbeda jika tanah pelapukan mengalami perubahan dan memperlihatkan karakteristik yang berbeda dengan tanah di sekitarnya, jarang penghuninya, dan terjadi perubahan lereng. Metode yang digunakan meliputi pengamatan tanah melalui deskripsi, identifikasi lereng, dan penggunaan tata guna lahan. Alat yang digunakan bor tangan sederhana, klinometer dan Global Positioning System (GPS) dan MapInfo. Hasil kajian menunjukkan bahwa geologi permukaan/tanah hasil proses pelapukan batuan dasar tidak menunjukkan perubahan, perbedaan tanah dengan tanah yang ada di sekitarnya. Kata kunci: Tapak RDE, Pedologi, Patahan.

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P-37 ANALISIS PENGARUH WAKTU KONSTRUKSI TERHADAP KELAYAKAN FINANSIAL PROYEK PLTN SMR DI INDONESIA DENGAN PENDEKATAN PROBABILISTIK

Nuryanti1, Suparman2, Sufiana Solihat3 1, 2, 3 Pusat Kajian Sistem Energi Nuklir (PKSEN) BATAN

Jl. Kuningan Barat, Mampang Prapatan, Jakarta 12710, Telp./Fax: (021)5204243 e-mail: [email protected]

ABSTRAK

ANALISIS PENGARUH WAKTU KONSTRUKSI TERHADAP KELAYAKAN FINANSIAL PROYEK PLTN SMR DI INDONESIA DENGAN PENDEKATAN PROBABILISTIK. Pusat Listrik Tenaga Nuklir (PLTN) jenis SMR (Small Medium Reactor) merupakan salah satu alternatif teknologi pembangkitan yang dapat diusulkan di wilayah Luar Jawa Bali (LJB), terkait dengan karakteristik jaringan kelistrikan dan amanat PP No 79 tahun 2014 tentang Kebijakan Energi Nasional. Pada proyek PLTN sangat dimungkinkan terhadap terjadinya sejumlah ketidakpastian. Hal ini tentu berpengaruh terhadap kelayakan finansial proyek. Salah satu variabel yang berpotensi menimbulkan ketidakpastian tersebut adalah waktu konstruksi. Variabel ini sangat berkorelasi dengan kelayakan finansial proyek terkait dengan nilai waktu uang. Oleh karena itu, penelitian ini bertujuan untuk mengkaji pengaruh dari waktu konstruksi terhadap kelayakan finansial PLTN SMR dengan pendekatan probabilistik. Analisis probabilistik dilakukan dengan teknik Monte Carlo yang mensimulasikan keterkaitan di antara variabel-variabel ketidakpastian dan dilihat pengaruhnya terhadap nilai indikator kelayakan finansial. Indikator kelayakan yang digunakan adalah nilai kini bersih (Nett Present Value – NPV) dan tingkat pengembalian internal (Internal Rate of Return – IRR). Hasil penelitian menunjukkan bahwa bertambahnya waktu konstruksi berakibat pada makin tidak layaknya proyek PLTN SMR sebagaimana tercermin pada semakin kecilnya nilai rata-rata NPV dan rata-rata IRR pada proyek. Probabilitas tertolaknya proyek pada ketiga waktu konstruksi yang disensitivitaskan masing-masing adalah sebesar 10% untuk waktu konstruksi 5 tahun, 40% pada waktu konstruksi 8 tahun dan 80% pada waktu konstruksi 10 tahun. Kata kunci: pendekatan probabilistik,waktu konstruksi,SMR,NPV, IRR

P-38 DETEKSI CACAT SAMPEL LAS MATERIAL SA533-B1 BEJANA TEKAN DENGAN METODA UJI TAK RUSAK

Mudi Haryanto, Sri Nitiswati, Andryansyah, Deswandri, Geni Rina

PTKRN-BATAN, KawasanPuspiptekSerpong, Gedung No. 80, Setu-Tangerang Selatan-15313 Email: [email protected]

ABSTRAK

DETEKSI CACATSAMPEL LAS MATERIAL SA533-B1 BEJANA TEKAN DENGAN METODE UJI TAK RUSAK. SA533-B1 adalah materialyang banyak digunakan pada meterial bejana tekan reaktor PWR. Umumnya pembuatan bejana tekan dilakukan dengan sambungan las. Untuk mengetahui kualitas pengelasan perlu dilakukan uji mekanik. Persyaratan sebagai benda uji mekanik adalah pada bagian sampel las dimana akan dibuat benda uji mekanik harus terbebas dari cacat.SampellasSA533-B1 di las denganproses SMAW. Tujuannya untuk mengetahui ada/tidaknya cacat las pada daerah sambungan yang di las, sehingga dapat diputuskan kemungkinan dapat/tidaknya sampel uji las dibuat untuk bahan benda uji mekanik.Metoda yang digunakan adalah uji tak rusak terdiri dari metode dye penetrant dan metode ultrasonik.Sampel las disediakan sebanyak 3 buah terdiri dari ukuran (132 x 130 x 12)mm sebanyak 1 buah dan ukuran (300 x 100 x 10)mm sebanyak 2 buah.Hasil dari pengujian dengan metoda dye penetrant dan metoda ultrasonik, pada sampel las nomor 1 ditemukan cacat porositi dipermukaan dan di dalam las, sampel las nomor 2 ditemukan cacat retak dipermukaan dan di dalam las, dan sampel nomor 3 ditemukan cacat lack of side wall fusion di dalam las. Semua cacat-cacat tersebut tidak merata sepanjang sambungan las,sehingga disimpulkan bahwa sampel las dapat dibuat benda uji mekanik pada daerah yang tidak ada cacatnya. Kata kunci : Sampel las SMAW, SA533-B1, bejana tekan, uji tak rusak, uji mekanik

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P-39 ANALISIS SPASIAL TATARUANG PROGRAM RDE DI KAWASAN PUSPIPTEK SERPONG

Heni Susiati1, Hadi Suntoko1, Sriyana1, Habib Subagio2

1 Pusat Kajian Sistem Energi Nuklir (PKSEN)-BATAN Jl. Kuningan Barat, Mampang Prapatan, Jakarta Selatan, 12710

2 Pusat Pemetaan dan Tataruang, Badan Informasi Geospasial (BIG) Jl. Raya Bogor, Jawa Barat


ABSTRAK ANALISIS SPASIAL TATARUANG PROGRAM RDE DI KAWASAN PUSPIPTEK, SERPONG. Kawasan Pusat Penelitian Ilmu Pengetahuan dan Teknologi (PUSPIPTEK) di Kecamatan Setu sebagai modal besar untuk menerapkan konsep pengembangan teknologi tinggi. Rencana pembangunan RDE di kawasan PUSPIPTEK menurut persyaratan BAPETEN, diperlukan evaluasi terhadap perkembangan dan kesesuaian terhadap tataruang kawasan PUSPIPTEK dan sekitarnya. Tujuan penelitian adalah melakukan evaluasi kondisi eksisting status polaruang dan kesesuaian terhadap rencana tataruang wilayah di Tangerang Selatan dengan wilayah studi pada radius 20 km dari pusat rencana pembangunan RDE. Metodologi yang digunakan dengan interpretasi data pengunaan lahan dari citra satelit. Kondisi wilayah Tangerang Selatan yang berkembang cukup pesat akan berpengaruh terhadap rencana program pembangunan RDE, khususnya terhadap perkembangan pemukiman. Hasil analisis polaruang sehubungan dengan rencana pembangunan RDE menunjukkan bahwa polaruang kawasan PUSPIPTEK dan sekitarnya dalam radius 20 km didominasi oleh pemukiman. Rencana pembangunan RDE di kawasan PUSPIPTEK sudah sesuai dengan RTRW Kota Tangerang Selatan. Namun demikian masih diperlukan program perencanaan pembangunan yang terpadu terkait dengan rencana penggunaan ruang kota Tangerang Selatan. Kata kunci: PUSPIPTEK, BAPETEN, RDE, tataruang, citra satelit, pemukiman

P-40 CLEARING HOUSE TEKNOLOGI NUKLIR BERBASIS STANDARDISASI: STUDI KASUS PADA ASSESSMENT TEKNOLOGI REAKTOR NUKLIR GENERASI IV

I Wayan Ngarayana, Sigit Santosa Pusat Standardisasi & Mutu Nuklir, Badan Tenaga Nuklir Nasional, Serpong 15314

ABSTRAK

CLEARING HOUSE TEKNOLOGI NUKLIR BERBASIS STANDARDISASI: STUDI KASUS PADA ASSESSMENT TEKNOLOGI REAKTOR NUKLIR GENERASI IV. Clearing house merupakan organisasi atau pengorganisasian yang memiliki peran penting dalam hal perlindungan dan juga memberikan informasi yang valid mengenai suatu teknologi kepada para stakeholder. Sampai dengan hasil kajian ini selesai ditulis, Indonesia belum memiliki skema baku dalam pelaksanaan Clearing House Teknologi Nuklir. Padahal teknologi nuklir sudah dimanfaatkan secara luas dalam berbagai bidang kehidupan, baik yang dihasilkan dari program penelitian dan pengembangan di dalam negeri, maupun yang didatangkan dari luar negeri. Salah satu contoh pelaksanaan Clearing House Teknologi Nuklir yang harus segera dihadapi adalah pelaksanaan clearance terhadap teknologi reaktor nuklir generasi IV yang sudah ditawarkan untuk dapat dibangun di Indonesia, yaitu Molten Salt Reactor yang berbahan bakar thorium. Kajian ini bertujuan untuk memformulasikan pelaksanaan clearing house yang selaras dengan konsep-konsep yang mungkin sudah dikembangkan serta diterapkan secara nasional dan/atau internasional sebagai instrumen dalam mekanisme clearance teknologi nuklir jenis baru. Dari hasil kajian yang telah dilakukan secara kualitatif melalui metode pengumpulan data secara secondary analysis dan focus group discussion, diperoleh kesimpulan bahwa kegiatan Clearing House Teknologi Nuklir pada dasarnya dapat didekati melalui perluasan implementasi standardisasi ketenaganukliran. Kata Kunci: Clearing House, Clearance Test, Standardisasi, Teknologi Nuklir

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P-41 STUDI DAMPAK PEMBUANGAN KONSENTRAT DESALINASI RO TERHADAP BIOTA PERAIRAN MANGGAR

Siti Alimah1, Heni Susiati1, June Mellawati2 1Pusat Kajian Sistem Energi Nuklir (PKSEN)-BATAN Jl. Kuningan Barat, Mp Prapatan, Jakarta

2Pusat Teknologi Keselamatan dan Metrologi Radiasi (PTKMR), Jakarta Selatan E-mail :[email protected]

ABSTRAK

STUDI DAMPAK PEMBUANGAN KONSENTRAT DESALINASI RO TERHADAP BIOTA PERAIRAN MANGGAR. Teknologi desalinasi MED-RO dengan pasokan energi dari PLTN dapat menjadi alternatif untuk memasok kebutuhan air bersih di Provinsi Kalimantan Timur. Buangan konsentrat desalinasi (brine) dapat memberikan dampak bagi kelangsungan hidup biota laut sekitar tapak PLTN karena salinitasnya tinggi dan mengandung aditif kimia yang digunakan saat pengolahan awal air umpan. Lingkungan dengan beragam biota laut dan sensitif terhadap buangan konsentrat desalinasi RO adalah mangrove dan terumbu karang, yang mana juga terdapat di perairan pesisir Manggar. Tujuan studi adalah mengetahui perkiraan dampak terhadap biota perairan pesisir Manggar jika dilakukan pembuangan konsentrat desalinasi RO. Peningkatan salinitas dapat mengurangi plankton, kerentanan krustacea dan larva invertebrata yang mengambang di air laut serta kerentanan ikan. Metode yang digunakan adalah pengumpulan data sekunder serta kajian literatur. Selanjutnya dilakukan analisis dan evaluasi data. Data sekunder diperoleh dari BLH daerah & pusat maupun Kementerian Kehutanan. Hasil studi menunjukkan bahwa berbagai aditif kimia yang digunakan dalam pengolahan awal air laut umpan desalinasi RO yang dapat tersisa di konsentrat jika tidak ada penanganan sebelum pembuangan yaitu sodium bisulfit, asam sulfat, sodium tripolifosfat, sodium hexametafosfat, EDTA dan asam sitrat. Sodium bisulfit dapat menyebabkan kematian biota di badan air wilayah pesisir, terutama ikan dan kepiting. Sedangkan asam sulfat dapat menurunkan pH sehingga mengurangi populasi ikan dan invertebrata. Limbah fosfat dari sodium tripolifosfat atau sodium hexametafosfat dapat menyebabkan kematian massal ikan dan keracunan kerang. Adanya EDTA dalam air laut dengan kesadahan tinggi akan menimbulkan efek racun dan kematian ikan. Asam sitrat di perairan laut hanya memberikan efek racun pada kepiting, alga dan protozoa. Namun, dampak pembuangan konsentrat di sekitar tapak tersebut dapat diminimalkan jika terdapat penanganan yang tepat sebelum konsentrat tersebut dibuang, yang dapat meliputi netralisasi, pengenceran dan pembuangan melalui pipa yang sangat panjang (50-300 m) serta dispersi dengan memperhatikan batimetri, gelombang, arus dan kedalaman laut. Kata kunci: dampak pembuangan, konsentrat desalinasi, RO, biota laut, perairan Manggar.

P-42 KANDUNGAN LOGAM BERAT DALAM AIR KALI PESANGGRAHAN DISEKITAR KAWASAN NUKLIR PASAR JUMAT (KNPJ)

Roza Indra L, Sri Widarti, Andung Nugroho, Miki Arian S

1 Pusat Teknologi Bahan Galian Nuklir-BATAN, Jl. Lebak Bulus Raya No.9, Jakarta email: [email protected]

ABSTRAK

KANDUNGAN LOGAM BERAT DALAM AIR KALI PESANGGRAHANDISEKITAR KAWASAN NUKLIR PASAR JUMAT (KNPJ). Untuk melihat adanya migrasi logam berat yang berasal dari kegiatan di Kawasan Nuklir Pasar Jumat (KNPJ) ke wilayah badan air Kali Pasanggrahan, maka perlu dilakukan pengamatan kandungan logam berat dan parameter lainnya. Pengamatan kandungan logam berat dalam air Kali Pasanggrahan telah dilakukan pada bulan September 2016. Contoh air diambil pada 7 titik pengambilan sampel.Kandungan logam berat diukur dengan menggunakan Spektrofotometer Serapan Atom (SSA).Penelitian ini bertujuan untuk mengetahui kandungan logam berat Cd, Pb, Cu, Zn, Co, Fe, Cr Dan Mn yang terdapat dalam air Kali Pasanggrahan tersebut, sehingga hasil yang diperoleh dapat memberikan informasi kepada masyarakat sekitar tentang kelayakan untuk kebutuhan sehari-hari. Parameter lainnya yang diuji adalah pengukuran suhu, pH, total disolved solid (TDS), Disolved Oxygen (DO) dan Biological Oxygen Demand (BOD).Metode yang digunakan untuk pengujian logam berat dengan metode spektrofotometri menggunakan Spektrofotmeter Serapan Atom (SSA). Sedangkan untuk pengukuran suhu, pH, TDS, DO dan BOD menggunakan DO meter. Dari pengujian dengan menggunakan menunjukkan bahwa kandungan logam Cd, Cu, Zn, Co, Cr dan Mn tidak terdeteksi (ttd), tapi terdeteksi kandungan logam Pb dan Fe dengan nilai tertinggi masing-masing sebesar 0,02 mg/l dan 0,012 mg/l. Namun nilai tersebut tidak melebihi batas ambang baku mutu Peraturan Menteri Negara Lingkungan Hidup Nomor 82 tahun 2001, dimana kandungan logam berat Pb yang diperbolehkan yaitu Pb sebesar 0,03 mg/L dan Fe sebesar 0,3 mg/l. Nilai pH, TDS berada dibawah nilai ambang baku mutu kategori kelas, DO berada di kategori kelas II dan BOD berada pada kategori kelas III . Kata kunci: logam berat, pH, Total Disolved solid, Disolved Oxygen, Biological Oxygen Demand

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P-43 PERFORMANCE EVALUATION OF ADHOC PROTOCOLS: AODV AND DSDV FOR MOBILE NODE REQUIREMENT USING NS-2

A. A. Waskita1, D. Andiwijayakusuma1, Deswandri1, Geni R. Sunaryo1

1 Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency, Kawasan Puspiptek Serpong, Tangerang 15310, Indonesia


ABSTRACT PERFORMANCE EVALUATION OF ADHOC PROTOCOLS: AODV AND DSDV FOR MOBILE NODE REQUIREMENT USING NS-2. A performance evaluation of adhoc routing protocol of AODV and DSDV have been performed using NS-2 software. This evaluation was conducted to know which protocol was performed better in a mobile condition. The performance of both protocols is determined by the factor of the higher rate of Packet Delivery Fraction and the lower rate of Average end-to-end delay in a different number of nodes, the packet size and the maximum speed of nodes. From the simulation result, AODV shows good performance compare to DSDV in a mobile node requirement. Keyword: performance evaluation, AODV, DSDV, mobile nodes, NS-2

P-44 CONVERSION OF CO2 TO HYDROCARBON SYNFUEL BY UTILIZING NUCLEAR HYDROGEN COGENERATION

Djati H Salimy, Syaiful Bakhri, Geni R Sunaryo

Center for Nuclear Reactor Technology and Safety (PTKRN) National Nuclear Energy Agency of Indonesia (BATAN)

Puspiptek Area SerpongTangerang Selatan 15310 Indonesia Email: [email protected]

ABSTRACT

CONVERSION OF CO2 TO HYDROCARBON SYNFUEL BY UTILIZING NUCLEAR HYDROGEN COGENERATION. A study of the utilization of hydrogen cogeneration with nuclear energy as a technology for the conversion of CO2 into synthetic liquid hydrocarbon fuels has been carried out. The aim of the study is to understand the conversion of CO2 and H2 into synthetic fuels, as well as the role of nuclear hydrogen cogeneration for the production of hydrogen and as a source of process heat energy. The method used is literature study based on the results of existing research. Conventionally, synthetic fuel production from coal is produced through coal gasification process, followed by reacting synthesis gas (mixture of CO and H2) in FT reactor to synthesis fuel. In this study, we studied the production of synthetic fuels with CO2 and H2 raw materials. CO2 comes from emissions of coal-fired plants, whereas H2 is produced by nuclear hydrogen cogeneration systems. The results show that compared to conventional processes, CO2 and H2-based processes supported by coal cogeneration systems provide significant advantages in terms of CO2 emissions. The process based on coal gasification and nuclear cogeneration, capable of reducing emissions by up to 75% and saving up to 40% of coal consumption. While the process based only on CO2 and nuclear hydrogen cogeneration (without coal gasification), teoretically can operate witout any CO2 emission at all. Even this process can captured and utilize CO2 emissions from coal fired plant, and use it as a raw material for the process. Keywords: cogeneration system, CO2 conversion, CO2 emission, synfuel

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P-45 PEMANTAUAN METEOROLOGI PADA CALON TAPAK PLTN DI DESA SEBAGIN PULAU BANGKA

Denissa Beauty Syahna1, Kurnia Anzhar1, Slamet Suryanto1,

1 Pusat Kajian Sistem Energi Nuklir, Jl. Kuningan Barat, Mampang Prapatan, Jakarta 12710 email: [email protected]

ABSTRAK

PEMANTAUAN METEOROLOGI PADA CALON TAPAK PLTN DI DESA SEBAGIN PULAU BANGKA. Dalam rangka penyiapan lokasi tapak PLTN telah dilakukan studi kelayakan tapak PLTN di Pulau Bangka. Kegiatan ini menghasilkan dua lokasi calon tapak yaitu, di Mentok Kabupaten Bangka Barat dan Sebagin Kabupaten Bangka Selatan. Salah satu aspek yang perlu dipantau dalam studi kelayakan tapak adalah aspek meteorologi. Pada makalah ini dibahas hasil pemantauan meteorologi di stasiun Sebagin. Pemantauan dilakukan secara terus-menerus selama 24 jam/hari dengan menggunakan peralatan Automatic Weather Station (AWS) yang terdiri dari data logger dan sensor. Data yang diperoleh diantaranya suhu, kelembapan relatif, radiasi matahari, curah hujan, arah angin dan kecepatan angin pada tahun 2016. Tujuan penulisan ini adalah menyajikan hasil dari pengolahan dan analisis data meteorologi di stasiun Sebagin pada tahun 2016. Metode yang digunakan meliputi pemantauan, pengolahan data dan analisis. Hasil pengolahan dan analisis data meteorologi memperoleh nilai maksimum suhu sebesar 32,9°C, radiasi matahari 1142 W/m2, curah hujan 30,9 mm dan kecepatan angin 0,5 – 2,1 m/s. Nilai ini menunjukkan bahwa kondisi meteorologi di calon tapak Sebagin tidak terdapat nilai ekstrim parameter meterologi. Kata kunci: meteorologi, tapak PLTN, Pulau Bangka, pemantauan, nilai ekstrim

P-46 DESAIN DASHBOARD UNTUK MENDUKUNG PROSES PENGAMBILAN KEPUTUSAN PEMASANGAN KAPASITOR DAYA PADA SALURAN 20 KV

DI SEKITAR WILAYAH PLTN

Rizki Firmansyah Setya Budi, Wiku Lulus Widodo PKSEN-BATAN, Kuningan Barat, Jakarta Selatan 12710

E,mail: [email protected]

ABSTRAK Desain Dashboard untuk Mendukung Proses Pengambilan Keputusan Pemasangan Kapasitor Daya pada Saluran 20 kV di Sekitar Wilayah PLTN. Pada penelitian sebelumnya telah dilakukan optimasi penentukan letak dan ukuran kapasitor daya untuk perbaikan tegangan saluran 20 kV di sekitar PLTN. Hasil penelitian tersebut berupa tabel aliran daya. Penyajian hasil dalam bentuk tabel mengakibatkan pengguna sulit memahami dan menyebabkan kesalahan dalam pengambilan keputusan. Permasalahan tersebut dapat diselesaikan dengan menggunakan dashboard. Tujuan penelitian ini adalah membuat dashboard dengan menggunakan desain yang baik agar pengguna dapat mengambil keputusan yang tepat. Penelitian ini dilakukan dengan cara: studi literatur, analisis penentuan KPI, analisis desain dashboard, dan pembuatan dashboard. Hasil penelitian menunjukkan bahwa dashboard telah di buat dengan menggunakan desain yang baik, dengan indikator telah mencakup key performance indicator dan prinsip visualisasi data. Penggunaan dashboard ini akan mempermudah pengguna dalam proses pengambilan keputusan pemasangan kapasitor daya pada saluran 20 kV di sekitar wilayah PLTN. Kata kunci: dashboard, pengambilan keputusan, kapasitor daya, saluran 20 kV, PLTN

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P-47 FAKTOR PENYEBAB PENUNDAAN KONSTRUKSI PLTN DI DUNIA SEBAGAI PEMBELAJARAN UNTUK PEMBANGUNAN PLTN DI INDONESIA

Dharu Dewi

PKSEN BATAN, Jalan Kuningan Barat, Mampang Prapatan, Jakarta Selatan email: [email protected]

ABSTRAK

FAKTOR PENYEBAB PENUNDAAN KONSTRUKSI PLTN DI DUNIA SEBAGAI PEMBELAJARAN UNTUK PEMBANGUNAN PLTN DI INDONESIA Sebagai pembelajaran untuk pembangunan Pembangkit Listrik Tenaga Nuklir (PLTN) di Indonesia, maka sangat penting dipelajari berbagai faktor penyebab penundaan konstruksi PLTN yang terjadi di beberapa negara sehingga jika Indonesia jadi membangun, pemilik PLTN dapat melakukan antisipasi dalam pelaksanaan proyek konstruksi PLTN. Banyak negara seperti Argentina, Brasil, Cina, Korea, Bulgaria, Romania, dan lain-lain telah mengalami risiko penundaan konstruksi. Tujuan studi adalah mengidentifikasi faktor – faktor penyebab utama penundaan proyek konstruksi PLTN. Faktor penyebab tersebut perlu dikaji dengan baik sehingga dapat menjadi pembelajaran bagi Indonesia dalam membangun PLTN. Metodologi yang digunakan adalah kajian literatur/pustaka dari beberapa negara yang diperoleh dari dokumen IAEA, dan jurnal terkait dengan penundaan konstruksi PLTN. Hasil studi menyimpulkan bahwa faktor penyebab penundaan konstruksi PLTN pada prinsipnya serupa di setiap negara tergantung pada kondisi negara/pemerintah.. Faktor penyebab tersebut berupa risiko akibat kinerja yang buruk dari kontraktor, owner, badan regulasi, masalah pendanaan, faktor cuaca dan kondisi negara/pemerintah. Jika Indonesia membangun PLTN, maka faktor – faktor penyebab penundaan konstruksi PLTN perlu dipertimbangkan, diidentifikasi dan diantisipasi serta panduan manajemen risiko konstruksi perlu disusun secara rinci agar tidak terjadi penundaan konstruksi PLTN. Kata kunci: penundaan, PLTN, pembelajaran, faktor penyebab, konstruksi, dunia

P-48 PEMANTAUAN GEMPA MIKRO DI CALON TAPAK PLTN MURIA JAWA TENGAH TAHUN 2015

Hajar Nimpuno Adi, Kurnia Anzhar

Pusat Kajian Sistem Energi Nuklir – BATAN, Jl. Kuningan Barat, Mampang Prapatan, Jakarta Selatan 12710

Telp./Fax.: 021 - 5204243 email: [email protected]

ABSTRAK

PEMANTAUAN GEMPA MIKRO DI CALON TAPAK PLTN MURIA JAWA TENGAH TAHUN 2015. Penyiapan data tapak merupakan kegiatan yang sangat penting dalam rangka persiapan pembangunan PLTN. Keakuratan data tapak merupakan faktor yang sangat penting sebagai masukan rancang bangun (dasar desain) PLTN yang berbasis keselamatan yang spesifik untuk setiap tapak. Data gempa mikro harus didukung oleh perangkat keras dan perangkat lunak dalam menangani akuisisi, pengolahan dan analisis. Tujuan kajian ini adalah untuk mengetahui kondisi kejadian gempa khususnya gempa mikro di sekitar calon tapak PLTN di wilayah Semenanjung Muria tahun 2015. Jaringan pemantauan gempa mikro di Semenanjung Muria terdapat di 8 (delapan) lokasi stasiun dan merekam data secara kontinyu. Pengolahan data hasil pemantauan gempa dilakukan dengan menggunakan software SEISAN. Hasilnya diperoleh bahwa selama tahun 2015 terekam sebanyak 690 kejadian gempa, dimana 50 kejadian gempa terletak dalam radius 150 Km dari tapak Ujung Blitar, 151 kejadian gempa dalam radius 150 Km – 500 Km, sedangkan gempa lainnya di luar radius 500 km. Kejadian gempa dalam radius 150 km memiliki rentang magnitudo antara 1,9 – 4,9 SR. Nilai percepatan tanah maksimum dari gempa terdekat pada tanggal 23 Oktober 2015 pukul 02:10:19 diperoleh sebesar 128,9 gal (0,129 G). Kata kunci: Gempa mikro, Tapak PLTN, Muria.

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P-49 STUDI LITERATUR “PENGUKURAN LAJU EMISI NEUTRON SECARA ABSOLUT DENGAN SISTEM MANGANESE SULPHATE BATH (MnSO4.H2O) DI CIAE, CMI, KRISS,

LNE-LNHB, LNMRI, NIST, NPL, dan VNIIM”

Nazaroh Pusat Teknologi Keselamatan dan Metrologi Radiasi,Kawasan Nuklir Pasar Jumát,Jakarta 12440


ABSTRAK STUDI LITERATUR “PENGUKURAN LAJU EMISI NEUTRON SECARA ABSOLUT DENGAN SISTEM MANGANESE SULPHATE BATH (MnSO4.H2O) DI CIAE, CMI, KRISS, LNE-LNHB, LNMRI, NIST, NPL, dan VNIIM”. Metrologi Neutron adalah ilmu pengetahuan yang mempelajari tentang pengukuran neutron, Dalam metrologi neutron, besaran fisika primer adalah laju emisi neutron dan fluens neutron. Laju emisi neutron, dN/dt adalah jumlah neutron yang dipancarkan dari suatu sumber. Pengukuran laju emisi neutron AmBe, yang diselenggarakan oleh BIPM, diikuti oleh 8 partisipan (CIAE, CMI, KRISS, LNE-LNHB, LNMRI, NIST, NPL dan VNIIM). Teknik manganese sulphate bath merupakan salah satu metode “Penentuan laju emisi neutron secara absolut dari sumber neutron radionuklida”. Prinsip dari metode MnSO4.H2O bath adalah neutron yang datang dan memasuki bath, ditermalkan dan ditangkap larutan MnSO4 oleh berbagai inti yang ada di dalamnya, dan dengan fraksi tertentu ditangkap oleh 55Mn. Laju disintegrasi absolut dari 56Mn yang dihasilkan diukur dengan sistem pencacah koinsidensi 4 atau metode lain dan dengan menggunakan persamaan Axton, dapat diperoleh laju emisi neutron yang datang”. Koreksi dilakukan terhadap neutron leakage, self capture dan fast neutron capture oleh sulfur dan oksigen menggunakan sistem kode transport MCNP (Monte Carlo N-particle)-ENDF/B-VI cross section. Berdasarkan uji 2, laju emisi neutron rata-rata dari ke 6 partisipan adalah : (2,438 ± 0,088).106 n/s. Kata kunci: laju emisi, Manganese sulphate bath, CIAE, CMI, KRISS, LNE-LNHB, LNMRI, NIST, NPL, VNIIM

P-50 KAJIAN IMPLEMENTASI PLTN DI INDONESIA: PEMBELAJARAN DARI NEGARA PENDATANG BARU

Sahala Maruli Lumbanraja, Rr. Arum Puni Rijanti Pusat Kajian Sistem Energi Nuklir-BATAN Jl. Kuningan Barat, Mampang Prapatan-Jakarta

e-mail: [email protected]

ABSTRAK KAJIAN IMPLEMENTASI PLTN DI INDONESIA: PEMBELAJARAN DARI NEGARA PENDATANG BARU. Pertumbuhan penduduk dan peningkatan gaya hidup masyarakat Indonesia berimplikasi pada pertumbuhan kebutuhan energi listrik. Ketersediaan dan keamaan energi listrik merupakan prasyarat pertumbuhan ekonomi dan kesejahteraan penduduk. PLTN merupakan sumber energi yang telah banyak dimanfaatkan beberapa negara di dunia dan telah berkontribusi hingga 11,5 % dari kebutuhan energi primer di dunia. Tujuan dari studi ini adalah untuk mempelajari berbagai kendala dari aspek politik, sosial, ekonomi, pendanaan, sumber daya manusia, dan periinan yang timbul selama implementasi pra proyek hingga pembangunan PLTN. Kajian dilakukan dengan studi pustaka dari beberapa kasus kegagalan dan keterlambatan proyek pembangunan PLTN di dunia dan beberapa proyek infrastruktur di Indonesia. Keterlambatan penyelesaian proyek dan kegagalan implementasi PLTN umumnya disebabkan oleh perubahan politik, penolakan masyarakat, krisis ekonomi, keterbatasan pendanaan, kendala perijinan, ketersediaan dan kemampuan SDM yang kurang, juga keterlambatan dan/atau ketidakjujuran vendor dan supplier untuk memasok kebutuhan proyek sesuai kontrak. Kegagalan implementasi PLTN di Indonesia disebabkan oleh kurangnya dukungan politik, sedangkan keterlambatan penyelesaian proyek terutama disebabkan oleh keterlambatan penyelesaian perijinan, dan pembebasan lahan. PLTN yang akan dipilih sebaiknya dari vendor dan supplier yang dapat dipercaya dan masih aktif beroperasi. Indonesia harus mempersiapkan semua infrastruktur secara baik dan cermat untuk mengurangi resiko yang mungkin timbul. Kata kunci: SDM, vendor, supplier, teknologi, resiko

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P-51 ANALISIS STABILITAS SISTEM KELISTRIKAN BATAM DENGAN PENAMBAHAN PEMBANGKIT LISTRIK TENAGA NUKLIR

Citra Candranurani, Arief Tris Yuliyanto, Elok Satiti A, Rizki Firmansyah S.B, Rr. Arum Puni Rijanti

Pusat Kajian Sistem Energi Nuklir (PKSEN) – BATAN Jl. Kuningan Barat, Mampang Prapatan, Jakarta Selatan 12710


ABSTRAK ANALISIS STABILITAS SISTEM KELISTRIKAN BATAM DENGAN PENAMBAHAN PEMBANGKIT LISTRIK TENAGA NUKLIR. Kebutuhan daya listrik di Pulau Batam dan sekitarnya semakin meningkat seiring dengan peningkatan wilayah industri yang semakin berkembang. Untuk menyediakan daya listrik yang handal diperlukan perencanaan sistem kelistrikan yang baik, meliputi perencanaan pembangkit, transmisi dan distribusi. Rencana pembangunan Pembangkit Listrik Tenaga Nuklir (PLTN) 200 MW di sistem kelistrikan Barelang (Batam, Rempang, Galang) di tahun 2026 memerlukan studi aliran daya dan analisis stabilitas sistem agar diketahui keandalan sistem sebelum dan setelah penambahan PLTN. Studi ini dilakukan dengan bantuan perangkat lunak ETAP (Electrical Transient Analysis Program) dengan mengolah data kelistrikan eksisting dan perencanaan di sistem Barelang dari PT. PLN Batam. Studi aliran daya dilakukan dalam kondisi tunak atau stabil dan kondisi beban puncak. Hasil studi aliran daya diperoleh gardu induk (GI) yang menjadi kandidat penyaluran PLTN ada 6 (enam) yaitu Sei Baloi, Batu Besar, Sei Harapan, Tj. Sengkuang, Sagulung, dan Nongsa. Sedangkan analisis stabilitas dilakukan dalam kondisi transien pada GI Panaran dengan penyaluran daya PLTN di GI Sagulung. Dari hasil simulasi stabilitas yang dilakukan, diketahui bahwa GI Panaran dalam sistem Barelang tetap dalam kondisi stabil meskipun terjadi pemadaman (trip) PLTN secara mendadak. Kata kunci: studi aliran daya, analisis stabilitas sistem, kondisi tunak, kondisi transien, PLTN, Batam

P-52 PEMBUATAN SUMBER RADIOISOTOP 137Cs SEBAGAI STANDAR KALIBRASI PADA SPEKTROMETER GAMMA

Aslina Br.Ginting, Yanlinastuti, Boybul, Arif Nugroho, Dian A, Gatot W,Hermawan

PTBBN-BATAN, Kawasan PUSPIPTEK Serpong Tangerang Selatan

ABSTRAK PEMBUATAN SUMBER RADIOISOTOP 137Cs SEBAGAI STANDAR KALIBRASI PADA SPEKTROMETER GAMMA. Di dalam hotcell Instalasi Radiometalurgi (IRM) banyak larutan bahan bakar nuklir hasil pengujian bum up yang dihasilkan. Larutan tersebut belum dapat dilimbahkan karena masih mengandung radioisotop 137Cs dan hasil fisi lainnya yang mempunyai waktu paroh panjang, tetapi larutan tersebut dapat digunakan sebagai sumber standar sekunder radioisotop 137Cs untuk keperluan analisis. Dalam melakukan analisis bahan bakar pasca iradiasi selalu menggunakan metode spektrometer- yang valid dan terkalibrasi. Standar yang digunakan untuk mengkalibrasi aktivitas spektrometer- adalah radioisotop 137Cs. Permasalahannya adalah kalibrasi spektrometer- tidak dapat dilakukan secara rutin karena tidak tersedianya sumber standar. Kebutuhan radioisotop 137Cs untuk litbang masih tergantung dari luar negeri. Oleh karena itu, pada penelitian ini, PTBBN dan PTKMR bertujuan untuk membuat sumber standar sekunder radioisotop 137Cs dengan aktivitas 10330 ± 411 Bq. Standar diperoleh dari larutan hasil pemisahan radionuklida 137Cs dalam PEB U3Si2/Al pasca iradiasi. Pemisahan 137Cs dilakukan dengan metode pengendapan menggunakan serbuk CsNO3 dan HClO4 sebagai carier. Hasil pemisahan diperoleh endapan 137CsClO4, kemudian dikeringkan dan ditimbang, untuk selanjutnya diukur besar aktivitasnya menggunakan spektrometer-. Endapan 137CsClO4 kering kemudian dilakukan pengkemasan menjadi sumber standar tertutup (shield source) dan disertifikasi oleh PTKMR sehingga diperoleh standar sekunder radioisotop 137Cs yang siap digunakan untuk mengkalibrasi aktivitas spektrometer-. Kata kunci: larutan PEB U3Si2-Al, 137Cs,kalibrasi, standar sekunder dan spektrometer-.

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P-53 IDENTIFIKASI ALTERASI BATUAN BERDASARKAN RASIO Th/U di TAPALANG, MAMUJU, SULAWESI BARAT

I Gde Sukadana, Frederikus Dian Indrastomo, Ngadenin Center for Nuclear Minerals Technology, Lebak Bulus Raya No. 09, Jakarta, 12440.


ABSTRAK IDENTIFIKASI ALTERASI BATUAN BERDASARKAN RASIO Th/U di TAPALANG, MAMUJU, SULAWESI BARAT. Kawasan Mamuju telah dipelajari secara intensif untuk eksplorasi uranium sejak 2013. Radioaktivitas yang tinggi hanya ditemukan pada sebaran batuan gunung api Adang. Daerah yang memiliki kadar thorium tinggi umumnya memiliki tingkat alterasi batuan yang tinggi dengan rasio Th/U sangat bervariasi. Penelitian bertujuan untuk mengetahui sebaran alterasi batuan dan korelasinya terhadap rasio Th/U di Daerah Tapalang. Penelitian dilakukan dengan pengukuran radioaktivitas, pengamatan alterasi, dan analisis XRF dan analisis mineragrafi untuk mengetahui tingkat alterasi. Penelitian berlokasi di Kecamatan Tapalang yang meliputi Desa Takandeang, Orobatu dan Pasa’bu Daerah secara umum tersusun oleh lava dan breksi tapalang dengan komposisi ponolitic dan foiditic, dan sebagian kecil tersusun atas batugamping formasi Mamuju. Produk alterasi tersebut menunjukkan bahwa daerah ini telah dipengaruhi oleh alterasi hidrothermal pada potassic zone. Rasio Th/U pada batuan lava Tapalang yang masih relatif segar memiliki nilai 3-30, dan batuan yang telah teralterasi memiliki nilai 30 hingga lebih dari 3000. Nilai tersebut dapat digunakan sebagai dasar dalam melakukan deleniasi daerah alterasi yang memiliki rasio Th/U tinggi (30 - >3000). Proses alterasi meningkatkan tingkat kelarutan uranium, sehingga kadar uranium telah mengalami depleted dan pengkonsentrasian thorium serta logam tanah jarang (REE) yang signifikan. Pengembangan eksplorasi thorium dapat difokuskan pada daerah alterasi lanjut, sedangkan eksplorasi uranium harus difokuskan pada daerah yang bersifat reduktif yang memungkinkan terbentuknya cebakan uranium. Kata kunci: Rasio Th/U, alterasi, thorium, uranium, Mamuju.

P-54 MOLTEN SALT REACTOR (MSR) DENGAN DAUR BAHAN BAKAR THORIUM

Erlan Dewita, Sriyana

Pusat Kajian Sistem Energi Nukir, Jakarta 12710


ABSTRAK MOLTEN SALT REACTOR (MSR) DENGAN DAUR BAHAN BAKAR THORIUM.Dalam rangka untuk keamanan dan kemandirian energi nuklir, saat ini potensi penggunaan thorium sebagai bahan bakar nuklir alternatif banyak mendapatkan perhatian dunia karena kelebihan yang dimiliki. Selain, sifat-sifat fisiknya yang unggul, juga kelimpahannya di alam yang 3-4 kali lebih tinggi dibanding uranium. Pada dasarnya, bahan bakar basis thorium sudah digunakan pada beberapa jenis reaktor sejak tahun 60 an, seperti: Molten Salt Reactor di USA, MSRE dengan daya 7,4 MWth, HTGR (Dragon di UK, Pach bottom dan Fort St Vrain di USA), serta PWR 100 MW dioperasikan di Shippingport tahun 1977-1982. Reaktor MSR dipertimbangkan sebagai reaktor yang berpotensi menggunakan daur bahan bakar thorium. Oleh karena itu, perlu dilakukan studi dengan tujuan untuk lebih mengetahui dan memahami kinerja reaktor MSR dengan daur bahan bakar thorium dimana sebagai langkah awal apabila Indonesia akan mengembangkan reaktor tersebut di masa mendatang. Metodologi yang digunakan adalah mengkaji dan menganalisis beberapa pustaka hasil penelitian dan pengalaman dari beberapa negara terkait reaktor MSR menggunakan daur bahan bakar thorium dan sistem keselamatan. Hasil studi menunjukkan bahwa reaktor MSR sesuai untuk menggunakan daur bahan bakar thorium karena bentuk cair sehingga dapat dilakukan reprosesing secara online danisotop protactinium (Pa-233) yang terbentuk dan merupakan racun neutron dapat dipindahkan dari teras reaktor. Selain dari sisi keselamatan MSR menggunakan daur bahan bakar thorium dapat meningkatkan sistem keselamatan karena tidak akan terbentuk hidrogen yang dapat mengakibatkan ledakan serta tidak akan terjadi kecelakaan pelelehan teras. Hasil studi diharapkan dapat digunakan untuk menambah pemahaman terkait teknologi reaktor MSR dengan daur bahan bakar thorium yang kedepan akan berguna apabila Indonesia akan mengembangkan reaktor tersebut. Kata Kunci: thorium, bahan bakar, siklus, reaktor, uranium

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P-55 KOREKSI VARIASI HARIAN UNTUK SURVEI GEOMAGNETIK DI DAERAH POTENSI URANIUM DAN THORIUM, MAMUJU SULAWESI BARAT Dwi Haryanto, Adhika Junara Karunianto

Pusat Teknologi Bahan Galian Nuklir-BATAN, Jl. Lebak Bulus Raya 9, Pasar Jumat, Jakarta 12440 Email: [email protected]

ABSTRAK

KOREKSI VARIASI HARIAN UNTUK SURVEI GEOMAGNETIK DI DAERAH POTENSI URANIUM DAN THORIUM, MAMUJU SULAWESI BARAT. Metode geomagnetik merupakan metode geofisika yang dapat dilakukan untuk menentukan suseptibilitas medium di lokasi pengukuran. Koreksi variasi harian digunakan untuk menghilangkan pengaruh medan magnet luar pada nilai medan magnet hasil pengukuran. Pengukuran metode geomagnetik dilakukan menggunakan dua alat ukur untuk mendapatkan hasil yang lebih baik. Alat pertama ditempatkan pada lokasi tertentu (base) dan diset untuk melakukan pengukuran secara otomatis dalam periode tertentu. Alat kedua digunakan untuk melakukan pengukuran di titik-titik yang telah ditentukan pada desain survei. Lokasi penelitian dilakukan di daerah Takandeang, Mamuju, Sulawesi Barat. Pengukuran di base ini dilakukan untuk memperoleh data variasi harian yang akan dikoreksikan pada data hasil pengukuran di lapangan (rover). Penelitian ini bertujuan untuk menentukan nilai variasi harian untuk lokasi yang berbeda. Metode penentuan koreksi base untuk lokasi base yang berbeda dapat diterapkan untuk memecahkan masalah perbedaan lokasi base terutama untuk survei dengan lokasi yang luas.Data yang digunakan untuk melakukan koreksi merupakan data yang diperoleh dari base 2 dan base lain (1 atau 3) pada hari yang sama. Koreksi dari base 1 ke base 2 diambil dari hasil pengukuran tanggal 16, 28, dan 29 September 2016. Koreksi dari base3 ke base 2 diambil dari hasil pengukuran tanggal 10, 28, dan 29 September 2016. Nilai koreksi untuk base 1 ke base 2 sebesar 72,50 nT. Nilai koreksi untuk base 3 ke base 2 sebesar 40,25 nT. Nilai ini diperoleh dari perhitungan dengan bantuan kurva linier. Nilai ini merupakan selisih nilai dari base 2 dengan base 1 dan 3. Kata kunci: geomagnetik, variasi harian, koreksi, base, Mamuju

P-56 ANALISA DATA GEOMAGNETIK: STUDI KASUS DI WILAYAH CALON TAPAK RDE PUSPITEK-SERPONG DAN SEKITARNYA

Adhika Junara Karunianto1, Dwi Haryanto1, Fakhri Muhammad2 1 Pusat Teknologi Bahan Galian Nuklir-BATAN

Jalan Lebak Bulus Raya No. 9 Pasar Jumat Jakarta Selatan 12440 2 Universitas Negeri Solo

Jalan Ir. Sutami No. 36 A Surakarta 57126 email: [email protected]

ABSTRAK

ANALISA DATA GEOMAGNETIK: STUDI KASUS DI WILAYAH CALON TAPAK RDE PUSPITEK-SERPONG DAN SEKITARNYA. Metoda geomagnetik adalah salah satu metoda geofisika yang paling tua dan sudah banyak diaplikasikan di dunia eksplorasi dan survei tapak. Data geomagnetik di dalam penelitian ini berupa data gradiomagnetik yang pengukurannnya dilakukan dengan dua sensor dalam 1 alat magnetometer dengan jarak pisah 1.3m sejajar secara horisontal dengan interval waktu sekitar 5 detik. Area pengambilan datanya dilakukan di area calon tapak Reaktor Daya Eksperimen (RDE) Daerah Puspitek Serpong dan sekitarnya. Tujuan dari penelitian ini adalah mendapatkan peta anomali geomagnetik daerah Puspitek Serpong dan sekitarnya. Cara pengolahannya menggunakan metoda analisa tingkat lanjut yaitu teknik reduksi ke kutub dan upward continuation. Metoda pengolahan ini akan menghilangkan medan gradiomagnetik yang disebabkan oleh objek magnetik di atau dekat permukaan yang dianggap sebagai gangguan atau noise. Hasil penelitian berupa peta anomali medan gradiomagnetik yang terdistribusi secara lateral dengan rentang nilainya sekitar -400-450 nT. Berdasarkan peta tersebut dapat diperoleh tiga zona anomali yaitu westpart, centerpart, eastpart yang penyebarannya relatif arah utara-selatan. Bagian barat disebut zona westpart dengan nilai anomali sekitar 100-250 nT. Bagian tengah disebut zona centerpart dengan nilai anomali sekitar 50-450 nT, sedangkan bagian timur disebut eastpart dengan nilai anomali sekitar 50-250 nT. Kata kunci: geomagnetik, gradiomagnetik, sensor, noise, anomaly

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P-57 STUDI PENGARUH TEMPERATUR BAHAN BAKAR PADA KRITIKALITAS REAKTOR HOMOGEN MENGGUNAKAN SCALE

Arif Isnaeni

BAPETEN, Jl. Gajah Mada no. 8, Jakarta Pusat 10120 email: [email protected]

ABSTRAK

STUDI PENGARUH TEMPERATUR BAHAN BAKAR PADA KRITIKALITAS REAKTOR HOMOGENMENGGUNAKAN SCALE.Aqueous Homogeneous Reactor (AHR) adalah suatu bentuk reaktor yang menggunakan bahan bakar larutan uranium yang homogen. AHR memiliki karakteristik unik yang membedakan dengan reaktor non daya konvensional. Bahan bakar reaktor berupa larutan homogen campuran garam uranium dan juga moderator yang berada di dalam satu bejana reaktor. Saat ini sebagian besar radioisotop diproduksi di reaktor riset dan reaktor produksi isotop (reaktor heterogen) dengan metode iradiasi target yang mengandung bahan fisil U-235 diperkaya. Sebenarnya radioisotop tersebut juga dihasilkan di dalam bahan bakar reaktor, tapi tidak diekstrak. Berbeda dengan reaktor homogen, dimana bahan bakar dan target menjadi satu kesatuan. Sehingga seluruh radioisotop yang diperlukan dapat di ekstrak dari larutan bahan bakar. Bahan bakar tersebut akan menjadi bahan bakar bekas sebagai limbah radioaktif. Secara umum terdapat banyak sekali aspek yang perlu dipertimbangkan dalam mendesain reaktor yang aman dan sesuai standar keselamatan. Penelitian ini bertujuan untuk mengetahui pengaruh dari temperatur terhadap kritikalitas dari reaktor AHR. Simulasi perubahan temperatur dan kerapatan larutan bahan bakar pada reaktor homogen menggunakan program SCALE dengan bahan bakar Uranil Nitrat yang dilarutkan dengan H2O menunjukkan reaktor memiliki koefisien reaktivitas negatif terhadap temperatur dan kerapatan larutan bahan bakar. Fitur ini merupakan salah satu kelebihan dari reaktor homogen, sifat ini meningkatkan keselamatan reaktor homogen (AHR). Kata kunci: Temperatur, Mo-99, uranil nitrat, reaktor homogen, SCALE

P-58 PEMBUATAN MIKROHIDRO UNTUK MENUNJANG KEGIATAN PENELITIAN DI KAWASAN INSTALASI NUKLIR KALAN, KALBAR

Slamet, Singgih Andy Nugroho, Ahmad Dayani, Eddy Soesanto

Pusat Tenologi Bahan Galian Nuklir – BATAN Jln. Lebak Bulus Raya No. 9, Pasar Jumat,

Jakarta Selatan - 12440 [email protected]

ABSTRAK

PEMBUATAN MIKROHIDRO UNTUK MENUNJANG KEGIATAN PENELITIAN DI KAWASAN INSTALASI NUKLIR, KALAN, KALBAR. Pembuatan Mikrohidro telah dilakukan di Kawasan Instalasi Nuklir, Kalan, Kalbar. Tujuan kegiatan ini adalah untuk memenuhi kebutuhan energi listrik di kawasan Instalasi Kalan dengan memanfaatkan sumberdaya air yang tersedia. Kebutuhan sumber daya listrik di kawasan ini dulu menggunakan generator listrik yang tidak bisa dioperasikan 24 jam. Mengingat adanya aliran air sungai yang terus menerus mengalir baik musim kemarau apalagi musim penghujan, maka dibuatlah mikrohidro sebagai salah satu pilihan dalam pemenuhan kebutuhan daya listrik. Tahapan prosesnya meliputi: kajian teknis, pemilihan sistem mikrohidro, pembangunan fisik, pembuatan rumah mikrohidro, instalasi turbin dan peralatan pendukungnya, ujicoba mikrohidro, dan instalasi jaringan listrik. Hasil yang didapat yaitu dihasilkannya listrik dengan kapasitas daya 5 KW yang mampu untuk memenuhi kebutuhan penerangan 24 jam tanpa henti di Kawasan Instalasi Bahan Galian Nuklir, Kalan, Kalbar. Kata kunci: Generator, kajian teknis, turbin, mikrohidro

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P-59 STUDI KETERSEDIAAN THORIUM UNTUK MENINGKATKAN KEAMANAN ENERGI NUKLIR

Abimanyu Bondan, Siti Alimah, Hadi Suntoko

Pusat Kajian Sistem Energi Nuklir-BATAN,Jl Kuningan Barat, Mampang Prapatan, Jakarta, 12710 email: [email protected]

ABSTRAK

STUDI KETERSEDIAAN THORIUM UNTUK MENINGKATKAN KEAMANAN ENERGI NUKLIR. Kajian ketersediaan thorium untuk meningkatkan penyediaan bahan bakar nuklir dilakukan. Saat ini sebagian besar reaktor nuklir di dunia telah menggunakan bahan bakar uranium. Penggunaan bahan bakar uranium secara terus menerus akan menyebabkan menurunnya sumberdaya uranium. Oleh karena itu, bahan bakar nuklir alternatif seperti thorium perlu dikembangkan. Thorium adalah bahan fertil (dapat biak), sehingga penggunaannya sebagai bahan bakar, thorium harus diubah menjadi bahan fisil terlebih dahulu, yakni bisa dicampur dengan bahan fisil seperti uranium (U-235) diperkaya, plutonium (Pu-239) atau U-233. Thorium jika menyerap neutron akan menjadi U-233 yang menghasilkan energi sehingga dapat digunakan sebagai bahan bakar nuklir. Tujuan studi ini adalah mengetahui ketersediaan thorium untuk meningkatkan penyediaan bahan bakar nukir. Metodologi yang digunakan adalah kajian literatur dan analisis. Hasil studi menunjukkan bahwa sumberdaya spekulatif thorium di Indonesia tersebar di daerah sabuk granit timah sebesar 133.668 ton pada tahun 2016. Ketersediaan sumber daya thorium dunia terbesar adalah di negara India sebesar 846,000 ton. Kata kunci: thorium, bahan fertil, uranium, bahan bakar nuklir, keamanan.

P-60 SISTEM MANAJEMEN DOSIS PADA PROSES DAUR ULANG ZAT RADIOAKTIF TERBUNGKUS CESIUM-137 YANG SUDAH TIDAK DIGUNAKAN

Suhaedi Muhammad1, Rr.Djarwanti,RPS2,Susyati3

1,3 Pusat Teknologi Keselamatan dan Metrologi Radiasi,Kawasan Nuklir Pasar Jumát,Jakarta 12440 2 Pusat Teknologi Radioisotop dan Radiofarmaka,Kawasan Nuklir Serpong, Serpong 15310


ABSTRAK SISTEM MANAJEMEN DOSIS PADA PROSES DAUR ULANG ZAT RADIOAKTIF TERBUNGKUS CESIUM-137 YANG SUDAH TIDAK DIGUNAKAN. Kebutuhan zat radioaktif terbungkus Cesium-137 untuk logging minyak maupun batu bara jumlahnya semakin tahun semakin bertambah. Selama ini pemenuhan kebutuhan zat radioaktif terbungkus Cesium-137 tersebut berasal dari produk impor dengan harga yang relatif cukup mahal. Dengan terbitnya Peraturan Pemerintah Nomor 61 Tahun 2013 tentang Pengelolaan Limbah Radioaktif membuka peluang secara legal bagi Badan Tenaga Nuklir Nasional (BATAN) untuk melakukan proses daur ulang terhadap zat radioaktif terbungkus yang tidak digunakan yang berasal dari penghasil limbah. Guna melindungi keselamatan dan kesehatan para pekerja radiasi yang terlibat dalam kegiatan proses daur ulang zat radioaktif terbungkus Cs-137 yang sudah tidak digunakan,pemegang izin (PI) harus memenuhi persyaratan proteksi dan keselamatan radiasi sebagaimana ditetapkan di dalam pasal 10 Peraturan Kepala Badan Pengawas Tenaga Nuklir (BAPETEN) Nomor 4 Tahun 2013 tentang Proteksi Dan Keselamatan Radiasi Dalam Pemanfaatan Tenaga Nuklir. Salah satu upaya yang dapat dilakukan oleh PI untuk memenuhi persyaratan tersebut adalah dengan menerapkan sistem manajemen dosis yang di dalamnya mencakup sumber potensi penerimaan dosis, kebijakan penerapan nilai batas dosis (NBD), kebijakan penerapan nilai pembatas dosis (NPD),manajemen penerimaan dosis, perkiraan besarnya dosis, evaluasi penerimaan dosis dan tindaklanjut penerimaan dosis. Melalui penerapan sistem manajemen dosis ini dapat diperkirakan besarnya dosis yang diterima oleh personil yang terlibat dalam kegiatan proses daur ulang zat radioaktif terbungkus Cs-137 yang sudah tidak digunakan. Dari sini dapat diketahui apakah ada personil yang menerima dosis melebihi NPD tapi masih kurang dari NBD atau ada personil yang menerima dosis melebihi NBD, sehingga dapat diketahui tindaklanjut apa yang harus dilakukan oleh PI. Kata kunci: sistem manajemen, dosis, sumber terbungkus,daur ulang

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P-61 ANALISIS DATA RADIOMETRI SEKTOR LEMAJUNG, KALAN, KALIMANTAN BARAT

Heri Syaeful, Suharji, Dhatu Kamajati

Pusat Teknologi Bahan Galian Nuklir, Badan Tenaga Nuklir Nasional Jl. Lebak Bulus Raya No. 9, Pasar Jumat, Jakarta Selatan


ABSTRAK ANALISIS DATA RADIOMETRI SEKTOR LEMAJUNG, KALAN, KALIMANTAN BARAT. Pemetaan radiometri merupakan salah satu metoda dalam eksplorasi uranium. Analisis data radiometri diperlukan untuk mendapatkan hubungan antara data hasil pengukuran di permukaan dengan data bijih uranium yang didapatkan dari pekerjaan pemboran. Hasil analisis diharapkan dapat jadi acuan dalam prediksi keberadaan bijih uranium di bawah permukaan berdasarkan data radiometri permukaan, dengan memperhatikan proses oksidasi, pelapukan, dan aliran air yang berlangsung setelah pembentukan bijih. Selain itu analisis juga bertujuan untuk menguji prosedur interpretasi log gamma (ILG). Metode yang dilakukan dalam analisis adalah pembuatan peta iso-kadar dan rasio kadar, dan mengkomparasi data permukaan dengan bawah permukaan dan data dosis radiasi dengan kadar unsur. Hasil analisis menunjukkan terdapat korelasi yang baik antara data bijih uranium di bawah permukaan dengan data kadar U, rasio U/K, dan rasio U/Th. Dalam rangka analisis kesesuaian aplikasi metoda ILG untuk mendapatkan kadar U dari data TC GR maka disimpulkan terdapat korelasi yang sangat baik antara kedua data tersebut. Kata kunci: radiometri, eksplorasi, uranium, gamma ray

P-62 KARAKTERISASI HASIL IMOBILISASI ZEOLIT YANG MENGANDUNG LIMBAH THORIUM

Gustri Nurliati1, Yuni K. Krisnandi2

1 Pusat Teknologi Limbah Radioaktif, Kawasan Puspiptek, Tangerang Selatan15310 2 Departemen Kimia, Universitas Indonesia, Depok


ABSTRAK KARAKTERISASI HASIL IMOBILISASI ZEOLIT YANG MENGANDUNG LIMBAH THORIUM.Saat ini terdapat limbah thorium di Pusat Teknologi Limbah Radioaktif (PTLR) yang telah dikondisioning dalam drum baja karbon 200 liter dan disimpan di tempat penyimpanan sementara limbah radioaktif. Limbah thorium ini berasal dari tanah yang tercemar larutan thorium dari pabrik kaos lampu PT. Tasuma Jaya. Thorium merupakan radionuklida pemancar alfa dan berumur paro panjang sehingga memerlukan pengelolaan dengan benar. Oleh karena itu perlu dilakukan imobilisasi limbah thorium dengan matriks tertentu untuk meminimalisasi potensi terlepasnya radioaktif ke lingkungan. Telah dilakukan imobilisasi zeolit yang mengandunglimbah thorium simulasi menggunakan matriks polimer epoksi. Zeolit yang mengandung limbah thorium dicampur dengan polimer epoksi dan dicetak dalam cetakan berbentuk silinder (D=29,5 mm, t=24,5 mm). Karakterisasi blok limbah ditentukan dengan mengukur densitas, kuat tekan dan laju pelindihan. Hasil pengukuran menunjukkan bahwa kondisi optimaluntuk blok limbah ZA2B dan NaZ adalah pada kandungan limbah 30% dengan densitas 1,0765 g/cm3, kuat tekan 58,05 MPa dan laju pelindihan 0,52 g/cm2.hari untuk blok limbah ZA2B dan densitas 1,0647 g/cm3, kuat tekan 74,88 MPa dan laju pelindihan 0,52 g/cm2.hariuntuk blok limbah NaZ. Kata kunci: imobilisasi, limbah thorium, resin epoksi, zeolit.

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P-63 PENENTUAN IN-HOUSE STANDARD LOGAM TANAH JARANG

Mutia Anggraini1, Samin2, Budi Yuli Ani1, Kurnia Setiawan W1

1Pusat Teknologi Bahan Galian Nuklir – Batan, Jl. Lebak Bulus Raya 9, Pasar Jum’at, Jakarta, 12440

2Pusat Sains dan Teknologi Akselerator – Batan, Caturtunggal, Depok, Sleman,Yogyakarta,55281 Email: [email protected]

A B S T R A K

PENENTUAN IN HOUSE STANDARD LOGAM TANAH JARANG. BATAN melalui Pusat Teknologi Bahan Galian Nuklir (PTBGN) telah menguasai teknologi pemisahan logam tanah jarang (LTJ) dari monasit. LTJ hidroksida yang dihasilkan oleh PTBGN merupakan bahan intermediate yang selanjutnya akan diproses menjadi oksida LTJ ataupun unsur LTJ sesuai peruntukan penggunaannya. Banyak industri yang mengandalkan kualitas produknya pada LTJ baik industri dengan teknologi sederhana sampai teknologi tinggi. LTJ yang dihasilkan harus memiliki kualitas produk yang baik dan harus memenuhi standar produk, oleh sebab itu diperlukan sertifikat yang menyatakan bahwa produk tersebut layak digunakan sebagai bahan acuan (In-House standard). Penelitian ini bertujuan untuk membuat In-House standard LTJ dengan metode sampling, pengayakan, uji kadar air, uji homogenisasi dengan metode uji Fisher, uji stabilitas, dan penentuan nilai in-house berdasarkan nilai ketidakpastian pengukuran. Sampel LTJ hidroksida yang berasal dari monasit dan diproduksi di PTBGN-BATAN telah lolos uji kadar air, homogenitas, dan stabilitas sehingga sampel tersebut dapat dijadikan in-house standard. Kata kunci: LTJ, In-House Standard, Monasit, Bahan Acuan

P-64 ASSESSMENT OF THE RADIOLOGICAL IMPACT OF THE WASTE TREATMENT FOR HID LAMPS CONTAINING Kr-85 AND Th-232

Moch Romli, Suhartono

PTLR – BATAN, Gd. 50 Kawasan PUSPIPTEK Serpong, Tangerang Selatan, 15310 email: [email protected]

ABSTRACT

ASSESSMENT OF THE RADIOLOGICAL IMPACT OF THE WASTE TREATMENT FOR HID LAMPS CONTAINING KR-85 AND TH-232. High-Intensity Discharge (HID) lamp technology utilizing a radioactive substance into its composition materials. Based on the recommendations of the regulatory body, HID lamps waste of the lighting company processed by Center for Radioactive Waste Technology (CRWT). Due HID lamps waste is new waste types accepted by CRWT which contain Kr-85 and Th-232, it is necessary to study the safety related processing will be done. For solid waste such as HID lamps, one type of processing that can be done is compaction. From the stages of activities in compaction processing, there are potential hazards of external and internal exposure. Dose contributions received by workers coming from external radiation, exposure originating from the broken lamps and the unbroken lamps, as well as internal radiation exposure from the inhalation. From the calculation, each worker earned a total radiation dose of 6.39 μSv for overall HID lamps waste received from PT. Panasonic Lighting Indonesia which is wasting in 2014. Keyword: HID lamp, waste, Kr-85, Th-232, safety related processing, dose, compaction

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P-65 KARAKTERISASI LIMBAH RADIOAKTIF CAIR DAN OPTIMASI PENGOLAHAN DENGAN PENUKAR ION

Ajrieh Setyawan1, Ivana Oktavianita2 1PTLR-BATAN, Gd. 50 Kawasan PUSPIPTEK Serpong, Tangerang Selatan

2Analisis Kimia, IPB-BOGOR email: [email protected]

ABSTRAK

KARAKTERISASI LIMBAH RADIOAKTIF CAIR DAN OPTIMASIPENGOLAHAN DENGAN PENUKAR ION. Alternatif pengolahan limbah radioaktif cair dilakukan dengan penukar ion. Tujuan dari percobaan ini untuk menentukan optimasi proses penukar ion dan kebutuhan resin yang digunakan untuk mengolah limbah radioaktif cair. Percobaan dengan melakukan karakterisasi limbah awal dilanjutkan menghitung kebutuhan resin untuk pengolahan dan optimasi waktu pengadukan. Dari hasil percobaan diperoleh karakterisasi logam awal Fe 67.30 mg/L, Mg 6.90 mg/L, Al 28.34 mg/L, dan Ca 49.26 mg/L dan hasil akhir proses penukar ion kondisi optimal pada pengadukan cepat 120 rpm selama 15 menit dan pengadukan lambat 50 rpm selama 2 jam dengan akhir logam terlarut tidak teridentifikasi. Kata Kunci : Penukar Ion, Jart Test

P-66 PENGAMBILAN LOGAM TANAH JARANG DALAM PASIR SENOTIM

Tri Handini, Sri Sukmajaya Pusat Sains Dan Teknologi Akselerator


ABSTRAK PENGAMBILAN LOGAM TANAH JARANG DALAM PASIR SENOTIM. Telah dilakukan proses pengambilan logam tanah jarang (LTJ) dalam pasir senotim. Pasir senotim yang mempunyai kandungan utama itrium di dijesti dengan asam sulfat. Pada proses pengambilan tanah jarang (LTJ) ini di lakukan dengan variasi perbandingan berat pasir : asam sulfat, ukuran butir pasir, waktu dijesti dan proses quenching. Untuk mengetahui kadar logam tanah jarang (LTJ) dilakukan analisis menggunakan XRF. Diperoleh hasil terbaik pada perbandingan pasir dan asam sulfat 1 : 2, ukuran butir 200 mesh, waktu dijesti 5 jam, quenching menggunakan air dan tanah jarang (LTJ) terambil = 93 % Kata kunci: logam tanah jarang, senotim

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O-27 RANCANG BANGUN OMNIWHEEL ROBOT SEBAGAI SASARAN TEMBAK DINAMIS

Kamaruddin1, Rafiuddin Syam2

1 Teknik Mesin, Jurusan Teknik Mesin Politeknik Negeri Fakfak, Papua Barat, 98612 2Teknik Mesin, Fakultas Teknik Universitas Hasanuddin Makassar, Makassar, 90245

email : [email protected]

ABSTRAK RANCANG BANGUN ROBOT PENGGERAK SASARAN TEMBAK. Perkembangan teknologi Automasi dan Mekatronika saat ini yang kian pesat menuntut manusia harus berpacu dengan waktu dimana dibutuhkan suatu alat yang dapat bekerja dengan efektif dan efisien sehingga memudahkan manusia dalam melakukan aktifitasnya. Untuk membuat Robot Penggerak Sasaran Tembak alternatif yang efektif dan efisien serta bekerja secara otomatis, sehingga sasaran yang akan ditembak dapat digerakkan kembali apabila terjadi proses pergantian sasaran dengan sistem yang bekerja secara otomatis sesuai keinginan. Prototipe dari omni wheels robot yang kami buat terdiri dari 4 buah roda dengan bentuk segi empat sama sisi dengan sudut masing-masing roda ke roda lainnya adalah sebesar 900 dari titik tengah. Kata kunci: sasaran tembak, omniwheel, mikrokontroler, kinematika dan dinamika

O-28 POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER

Ruslan Bauna, Rafiuddin Syam, Hairul Arsyad, Amiruddin DepartemenTeknik Mesin, Fakultas Teknik Mesin, Universitas Hasanuddin

(email: [email protected])

ABSTRAK POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER.Untuk budidaya ayam broiler dengan populasi minimal 3.000 ekor pada umur 30-57 hari membutuhkan usaha pengangkutan pakan di atas kandang minimal 11487 kg. Perencanaan ini bertujuan menghasilkan pemegang pakan (end effector) robot yang mampu memegang pakan dalam karung dengan baik dan efektif serta meletakkan pada tempat yang telah ditentukan secara mandiri dengan menggunakan sistem kontrol. Proses desain pemegang pakanrobot diawali dengan pengambilan data dimensi pakan dalam karung yang akan dipegang oleh end effector robot dan diperoleh dimensi karung pakan 25 cm x 55 cm x 65 cm dengan massa 50 kg, selanjutnya dilakukan proses desain konstruksi end effector yang cocok untuk memegang pakan dengan baik menggunakan program desain tiga dimensi SolidWorks2010. Proses berikutnyaadalah mendesain sistem mekanik pemegang pakan agar gerakan yang dihasilkan sesuai dengan yang diharapkan sehingga tidak merusak wadah pakan dan tetap tercengkram dengan baik saat proses pengangkutan pakan berlangsung. Kemudian, pembuatan rangkaian elektronik dan sistemkendalimenggunakan mikrokontroler Arduino Mega. Setelah dilakukan perakitan semua kompenen sistem mekanik dan pemasangan sistem elektronik dan kontrol, kemudian dilakukan pengujian untuk mengevaluasi kinerja pemegang pakan robot yang telah dibuat. Dari hasil pengujian memperlihatkan bahwa semua komponen bekerja dengan baik sesuai perencanaan, khususnya kecepatan dan percepatan gerak pemegang pakan sehingga mampu memegang dan meletakkan pakan dengan baik dan efektif tanpa merusak wadah pakan. Dengandemikian robot manipulator dapat mengangkat dan membawa pakan secara secara efektif. Kata kunci: mekanisme, pemindah, arduino, kontrol, desain, end effector.

64

The 4th International Symposium on Smart Material and Mechanics (ISSMM)

2017

65

P-67 SHOULD A TERRITORY BE HOME TO THE NUCLEAR WASTE DUMP? STUDY CASE: SOUTH AUSTRALIA

G. Tanjung1, S. Bakhri2, S.L. Butar-butar2

1Indonesian Islamic Society of South Australia, 5 Falco Court, Flagstaff Hill 5159, Australia 2National Nuclear Energy Agency of Indonesia, Puspiptek Area, Serpong 15310, Indonesia


ABSTRACT On the 19th of March 2015, the South Australian Governor, Hieu Van Le, has announced an important statement in establishing a nuclear fuel cycle royal commission. The royal commission’s main objective is to examine the possibility of South Australia’s future role in the nuclear industry. This announcement has triggered a public debate among South Australia’s communities. Learning from the South Australia’s current situation, this paper describes general guidelines on how to safely and carefully set up nuclear waste depository and disposal facilities in a territory. These guidelines provides many recommendations that involve all stakeholders including the government, the industry and the whole community. In conclusions, an accountable and transparent process which involves all main stakeholders is a crucial key in setting up safe storage and disposal nuclear waste facilities in a territory. Keyword: nuclear waste, depository and disposal facilities, guidelines, South Australia

P-68 AN ANALYSIS OF RADIATION PENETRATION THROUGH THE U-SHAPED CAST CONCRETE JOINTS OF CONCRETE SHIELDING IN THE MULTIPURPOSE GAMMA

IRRADIATOR OF BATAN

Tanti Ardiyati, Bang Rozali, Kasmudin Center for Nuclear Facilities Engineering, National Nuclear Energy Agency of Indonesia (BATAN), Tangerang

Selatan, Indonesia

ABSTRACT An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased. Keywords: radiation penetration, cast concrete joints, multipurpose gamma irradiator.

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P-69 REQUIREMENTS ANALYSIS FOR AUXILIARY POWER OF APR 1400 NPP ON SYSTEMS ENGINEERING APPROACH

*M. G. Shahinoor Islam, Raouf M. Elfaramawy, Jung Jae-cheon, & Lim Hak-kyu

Dept. of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan, Korea *Corresponding author: [email protected]

ABSTRACT

The main objectives of this paper is to present the systems engineering approach for configuration management of Auxiliary Power of APR 1400 NPP by using EA (enterprise Architecture) Systems Modelling Language (SysML) Software. Configuration Management ensures that product functional, performance, and physical characteristics are properly identified, documented, validated, and verified to establish product integrity that the documentation against which systems or components produced are known. The review of regulation and licensing criteria for related to 4.16 kV class 1E Auxiliary Power of APR 1400 is the baseline for conducting Configuration Management. Requirements documents review and extract is based on tracking each single SHALL and/or SHOULD statements through decomposition and traceability process. The individual requirements are elicited from the upstream documents (regulatory body) and downstream documents (utility documents) and modelled in the EA platform. So, this design documents process and model by systems engineering approach can be helpful to understand easily the licensing procedures and regulations of 4.16 kV class 1E of APR 1400 NPP. Key Words: Requirement Analysis, APR1400 NPP, EA SysML Software, Configuration Management.

P-70 STATIC, DYNAMIC, AND FATIGUE ANALYSIS OF THE MECHANICAL SYSTEM OF ULTRASONIC SCANNER FOR INSERVICE INSPECTION OF RESEARCH REACTORS

Muhammad Awwaluddin, Kristedjo K., Khairul Handono, Ahmad H.

Center for Nuclear Facilities Engineering, National Nuclear Energy Agency, PUSPIPTEK Area, South Tangerang, Indonesia 15310

ABSTRACT

This analysis is conducted to determine the effects of static and dynamic loads of the structure of mechanical system of Ultrasonic Scanner i.e., arm, column, and connection systems for inservice inspection of research reactors. The analysis is performed using the finite element method with 520 N static load. The correction factor of dynamic loads used is the Gerber mean stress correction (stress life). The results of the analysis show that the value of maximum equivalent von Mises stress is 1.3698E8 Pa for static loading and value of the maximum equivalent alternating stress is 1.4758E7 Pa for dynamic loading. These values are below the upper limit allowed according to ASTM A240 standards i.e. 2.05E8 Pa. The result analysis of fatigue life cycle are at least 1E6 cycle, so it can be concluded that the structure is in the high life cycle category. Keywords: Ultrasonic Scanner, Research Reactor, Mechanical System, Fatigue, Static.

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P-71 THE MECHATRONIC SYSTEM DESIGN OF ULTRASONIC SCANNER FOR INSERVICE INSPECTION OF RESEARCH REACTOR

Khairul Handono, Kristedjo K., M. Awwaluddin and Ihsan Shobary.

Center for Nuclear Facilities Engineering, National Nuclear Energy Agency, PUSPIPTEK Area, South Tangerang, Indonesia 15310

ABSTRACT

The mechatronic system design of ultrasonic scanner for inservices inspection of Research Reactor has been conducted. The requirement designed must be reliable operated, safety to personnel and equipments, ease of maintenance and operation, protection of equipment mechanically, interchangeability of equipments and addition of the several model of probe immersion ultrasonic tranducer. In order to achieve the above goals and obtain the desired results, a mechatronic design based on mechanical and electronic practical experiences will be needed. In this paper consist of the mechanical design and the system mechanical movement using stepper motor control. The criteria and the methods of designs of mechanical and electronic equipments of the system have been discussed and investigated. A mechanical and instrumentation control system drawing and requirement of design will be presented as the outcome of the design. The designed of mechanical system is consequently simulated by solidwork software. The intention of the above research is to create solutions in different ways of inservice inspection of integrity of Reactor. Keywords : mechatronic, ultrasonic scanner, inservice inspection and research

P-72 NEUTRON DOSE RATE ANALYSIS ON HTGR-10 REACTOR USING MONTE CARLO CODE

Suwoto, H. Adrial, A. Hamzah, Zuhair, S. Bakhri, G. R. Sunaryo Center for Nuclear Reactor Technology and Safety – National Nuclear Energy Agency of Indonesia (BATAN),

PUPSPIPTEK Complex, Office Buliding No. 80, Serpong, Tangerang Selatan 15310, Indonesia, Telp. (021)756-0912, Fax. (021)756-0913,

E-mail: [email protected]

ABSTRACT Abstract. The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman's Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards. Keywords: HTGR-10, neutron dose rate, MCNP5v1.6, ICRP-74.

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P-73 EVALUATION OF THE AP1000 DELAYED NEUTRON PAREMETERS USING MCNP6

T.M. Sembiring, J. Susilo, S. Pinem

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Gd. No. 80 Serpong , Tangerang Selatan 15310 Indonesia


ABSTRACT The MCNP6 code contains numerous features, one of those is to determine the delayed neutron parameters. The accuracy of calculated delayed neutron parameters affect the accuracy of transient or dynamic condition. The objective of this paper is to determine the delayed neutron parameters of the advance PWR reactor, AP1000, using MCNP6 code with the recent ENDF/B evaluated nuclear data file ENDF/B-VII.1. The MCNP6 calculation results shows that the maximum difference occurred in the βi and λi parameters are 38.30% and 45.63%, respectively. The superiority of MCNP6 can be seen in the change of prompt neutron life time (ℓ) parameters that cannot be obtained by the deterministic code, so it can be used in the sensitivity analysis of the delayed neutron parameters. Based on this research work, the accident analysis of the AP1000 reactor use the effective delayed neutron fraction (βeff) of 0.0051 and the prompt neutron life time (ℓ) of 19.5 μs for the first cycle. Keywords: PWR, AP1000, delayed neutron parameters, MCNP6, ENDF/B-VII.1

P-74 THE CHANGE OF RADIAL POWER FACTOR DISTRIBUTION DUE TO RCCA INSERTION AT THE FIRST CYCLE CORE OF AP1000

J Susilo, L Suparlina, Deswandri, G R Sunaryo Center for Nuclear Reactor Technology and Safety

PUSPIPTEK Complex O.B. No.80, Setu, Tangerang Selatan, 15310 Email: [email protected]

ABSTRACT

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumed approximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion. The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe. Keywords: Power Factor, RCCA, AP1000, SRAC2006

69

P-75 PRELIMINARY STUDY FOR ALTERNATIVE CONCEPTUAL CORE DESIGN OF THE MTR RESEARCH REACTOR

Tukiran S., Surian P., Tagor MS, M. Subekti, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety, BATAN, Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan, 15310 Indonesia


ABSTRACT The utilization of the research reactor is increasingly widespread, especially for radioisotope production and testing of advanced materials and preference to use a compact core. The reactor core design has been determined on the maximum thermal flux in the middle of the core per MW. BATAN has designed several alternative research reactor cores. The purpose of this research reactor is to obtain the optimum reactor core configurations with the criteria to have a thermal neutron flux in the centre of the core with minimum of 1.0x1015 n/cm2 s. Power level of the research reactor is 60 MWth with U9Mo/Al fuel 85 cm of height. Design of plate-type fuels with a higher core results in the heat transfer to the coolant optimal. All 16 fuel assemblies and the 4 control rods are inserted into the core for this reactor. The core design calculations were carried out with the WIMSD-5B and BATAN-FUEL codes. Conceptual design calculation results show that the core configuration with 5 × 5 grids, all the fresh fuel, fuel loading of 470 g, a D2O reflector, a maximum thermal neutron flux in the central core is 1.09 x 1015 n/ cm2s and the cycle length is 33 days. The reactor core design is the most optimal for MTR type. For the equilibrium core, a fuel loading of 600 g results in the maximum thermal flux of 1.07x1015

n/cm2s and the two safety rods should be used in the core. Keywords: conceptual design, U9Mo/Al fuel, research reactor, WIMSD-5B code , BATAN-FUEL code.

P-76 COOLING PERFORMANCE ANALYSIS OF THE PRIMARY COOLING SYSTEM REACTOR TRIGA-2000 BANDUNG

I.D. Irianto, S. Dibyo, S. Bakhri, G.R. Sunaryo

Center for Nuclear Reactor Technology and Safety National Nuclear Energy Agency of Indonesia

Puspiptek Area, Building 80, Serpong,Tangerang 15310, Indonesia Email: [email protected]

ABSTRACT The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C. Keywords: Cooling System, the primary pump, mass flow rate, temperature.

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P-77 ANALYSIS OF HELIUM PURIFICATION SYSTEM CAPABILITY DURING WATER INGRESS ACCIDENT IN RDE

Sriyono, Rahayu Kusmastuti, Syaiful Bakhri, Geni Rina Sunaryo

Puspiptek Area Building 80, Serpong, South Tangerang City, 15310, INDONESIA Email: [email protected]

ABSTRACT

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident. Keywords : water ingress, accident, purification, HPS, capability, RDE

P-78 ANALYSIS OF RADIATION SAFETY FOR SMALL MODULAR REACTOR (SMR) ON PWR-100 MWE TYPE

P. M. Udiyani, I Husnayani, Deswandri, and G. R. Sunaryo

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia-BATAN Serpong-Tangerang Selatan, Indonesia


ABSTRACT Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

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P-79 MASTER LOGIC DIAGRAM: AN APPROACH TO IDENTIFY INITIATING EVENTS OF HTGRS

J H Purba

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek, Serpong, Tangerang Selatan – Banten, Indonesia


ABSTRACT Initiating events of a nuclear power plant being evaluated need to be firstly identified prior to applying probabilistic safety assessment on that plant. Various types of master logic diagrams (MLDs) have been proposed for searching initiating events of the next generation of nuclear power plants, which have limited data and operating experiences. Those MLDs are different in the number of steps or levels and different in the basis for developing them. This study proposed another type of MLD approach to find high temperature gas cooled reactor (HTGR) initiating events. It consists of five functional steps starting from the top event representing the final objective of the safety functions to the basic event representing the goal of the MLD development, which is an initiating event. The application of the proposed approach to search for two HTGR initiating events, i.e. power turbine generator trip and loss of offsite power, is provided. The results confirmed that the proposed MLD is feasible for finding HTGR initiating events. Keywords: Master logic diagram, initiating events, high temperature gas cooled reactor, probabilistic safety assessment.

P-80 MAIN STEAM LINE BREAK ACCIDENT SIMULATION OF APR1400 USING THE MODEL

OF ATLAS FACILITY

A S Ekariansyah, Deswandri, Geni R. Sunaryo Center for reactor nuclear safety and technology (BATAN), Puspiptek area, Setu, Tangerang Selatan


ABSTRACT A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptions for the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model. Keywords: main steam line break accident, ATLAS facility, APR1400, best-estimate, RELAP5

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P-81 ANALYSIS ON THE ROLE OF RSG-GAS POOL COOLING SYSTEM DURING PARTIAL

LOSS OF HEAT SINK ACCIDENT

Susyadi, Endiah P. H, Sukmanto D, Andi S. E, Hendro T. Syaiful B, Geni R.S. Center for Nuclear Reactor Safety and Technology -BATAN

Kawasan PUSPIPTEK Gedung 80, Setu, Tangerang Selatan -15310 Email: [email protected]

ABSTRACT

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result also reveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram. Keywords: RSG-GAS, RELAP5, pool cooling system, loss of heat sink, transient without scram

P-82 STUDY RELAP5 HELIUM PROPERTIES FOR HTGR THERMAL HYDRAULIC ANALYSIS

Surip Widodo, Anis Rohanda, Muhammad Subekti, Topan Setiadipura, Syaiful Bakhri, Geni Rina S.

Center for Nuclear Reactor Technology and Safety, BATAN, Puspiptek Area Building No 80, Serpong, Tangerang Selatan 15310, Indonesia

e-mail: [email protected]

ABSTRACT The system codes non-specific for HTGR such as RELAP5 has been utilized for HTGR thermal hydraulic analysis even helium gas property is not based on KTA 3102.1. However, those RELAP5 applications for HTGR above are merely based on the assumption that RELAP5 helium properties are comparable to the helium properties in the KTA 3102.1. Therefore, the study for comparing the helium properties used in RELAP5 and the helium properties in KTA 3102.1 is required. The objective of this paper is to study the appropriateness’ helium properties in RELAP5 code for high temperature gas reactor (HTGR) thermal hydraulic analysis. There has been an inclined interest in the scientific community in the study of the application RELAP5 for HTGR thermal hydraulic analysis. The KTA 3102.1 provides the helium properties that are the most commonly use for the HTGR thermal hydraulic analysis. For this study, the RELAP5 helium properties are compared with the helium properties in KTA 3102.1. The comparison results showed that the RELAP5 helium properties are satisfactory for the HTGR thermal hydraulic analysis. Keywords: Helium, KTA 3102.1, RELAP5, HTGR Thermal Hydraulic.

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P-83 STEADY STATE TEMPERATURE DISTRIBUTION INVESTIGATION OF HTR CORE

Sudarmono, Suwoto, Syaiful Bakhri, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety, Bld 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia


ABSTRACT Reactor operation safety is highly related to fuel temperature parameter in the reactor core. To maintain the fuel in good integrity (no crack or melting), fuel temperature should be continuously within licensing criteria. Fuel temperature parameter is one of safety parameters in nuclear reactor operation. Fuel temperature is determined by local heat flux. High heat flux generation will cause a change in heat equilibrium and consequently result a change in fuel temperature. In the present study, steady state temperature distribution investigations of pebble bed for HTR-10 have been performed. The calculation is performed by using VSOP’94 code to do the calculation by which the core is divided into 50 components to represent the positions of various material compositions and to model fuel into 5 layers. The analysis is carried out based on the input data of reactor parameters, core specification, and fuel specifications as well as layer data of HTR-10. The results of this simulation in HTR-10 core at steady state indicated that the maximum temperature is 969.9C at the solid material and 1031.3 C in the fuel at the steady state condition. These temperatures are much lower than that of the maximum safety margin of fuel pebble, i.e. 1600C, therefore, it can be concluded that TRISO fuel is able to contain all radioactive fission products. Keywords: HTR-10 core, TRISO fuel temperature, solid material temperature, steady state

P-84 DEVELOPMENT A COMPUTER CODES TO COUPLE PWR-GALE OUTPUT AND

PC-CREAM INPUT

S Kuntjoro, M Budi Setiawan, Nursinta Adi W, Deswandri, G R Sunaryo Puspiptek Complex, Building no. 80, Serpong, Tangerang Selatan 15310 Indonesia


ABSTRACT Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats. Often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data, so that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format

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P-85 EVALUATION ON CREEP PROPERTIES OF TYPE 316SS SERIES

Sri Nitiswati, Sudarno, Andryansyah, Deswandri, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety - National Nuclear Energy Agency of Indonesia, Puspiptek Area Building 80, Serpong, South Tangerang City, 15313, INDONESIA


ABSTRACT Evaluation on creep properties of type 316 stainless steel (SS) series has been done. Type of 316SS series is a candidate material for major structural components of high temperature reactor such as liquid metal reactor (LMR) because it has a good mechanical property in high temperature, compatibility with the coolant sodium, and adequate welding ability. The objective of this research is to obtain and compare creep properties between type 316SS and type 316LN SS materials. The method use by conducted creep rupture tests in ranges temperature of 600°C, 700°C and 800°C and under a constant applied stress level of 200 MPa (393 N). Research result obtained that time to rupture of type 316SS less than that of type 316LN SS. The values of extension (mm), creep strain (mm/mm), creep strain rate (mm/mm/hrs), and rupture elongation (%) of type 316SS are higher than that of type 316LN SS for the same temperatures and applied constant stress level. It is concluded that type 316LN SS has a good creep resistance than that of type 316SS. Keywords: Creep properties, type 316SS series, high temperature reactor

P-86 ANALYSIS OF JKT01 NEUTRON FLUX DETECTOR MEASUREMENTS IN RSG-GAS REACTOR USING LabVIEW

Rokhmadi1, Agus Nur Rachman1, Sujarwono2, Taswanda Taryo1, Geni Rina Sunaryo1

1 Center for Nuclear Reactor Technology and Safety Kawasan PUSPIPTEK Gd. 80 SerpongTangerang Selatan 15310

2 Center for Multipurpose Reactor – BATAN Kawasan PUSPIPTEK Gd. 80 SerpongTangerang Selatan 15310


ABSTRACT The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It,istherefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEWwas developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEWavailable in the PC. By using this simulation program, it is successfulto perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation resultsshowed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, , 1.5x, , 1.7x and 2.0x. Keywords : LabVIEW, Neutron Flux, Detector, Reactor

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P-87 NEUTRON FLUENCE AND DPA RATE ANALYSIS IN PEBBLE-BED HTR REACTOR VESSEL USING MCNP

Amir Hamzah, Suwoto, Anis Rohanda, Hery Adrial, Syaiful Bakhri and Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety - BATAN email: [email protected]

ABSTRACT

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82 x 108 n/cm2/s and 1.79 x 108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4 x 1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0 x 1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years. Keywords: Pebble-bed HTR reactor, fast neutron fluence, radiation damage, reactor vessel.

P-88 DETERMINING COOLANT FLOW RATE DISTRIBUTION IN THE FUEL-MODIFIED TRIGA

PLATE REACTOR

Endiah Puji Hastuti, Surip Widodo, M. Darwis Isnaini, Geni Rina S., Syaiful B. Center for Nuclear Reactor Technology and Safety, Bld 80 PUSPIPTEK Area, Serpong, South Tangerang,

15310, Indonesia Email: [email protected]

ABSTRACT

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔT onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35 C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively. Keywords: CAUDVAP code, COOLODN code, flow distribution, modified reactor, TRIGA plate

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P-89 PROBABILISTIC ANALYSIS ON THE FAILURE OF REACTIVITY CONTROL FOR THE PWR

D T Sony Tjahyani, Deswandri, G R Sunaryo

Center for Nuclear Reactor Technology and Safety, BATAN Puspiptek Complex, Building no. 80, Serpong, Tangerang Selatan 15310, Indonesia


ABSTRACT The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria. Keywords: Fundamental safety function, Probabilistic, Safety analysis, Reactivity control, PWR

P-90 ULTRASONIC NON-DESTRUCTIVE PREDICTION OF SPOT WELDING SHEAR

STRENGTH

R. Himawan, M. Haryanto, R.M. Subekti, and G.R. Sunaryo Center for Nuclear Reactor Technology and Safety, BATAN, Puspiptek Area, Building No. 80, Tangerang,

15310 Indonesia


ABSTRACT To enhance a corrosion resistant of ferritic steel in reactor pressure vessel, stainless steel was used as a cladding. Bonding process between these two steels may result a inhomogenity either sub-clad crack or un-joined part. To ensure the integrity, effective inspection method is needed for this purpose. Therefore, in this study, an experiment of ultrasonic test for inspection of two bonding plate was performed. The objective of this study is to develop an effective method in predicting the shear fracture load of the join. For simplicity, these joined was modelled with two plate of stainless steel with spot welding. Ultrasonic tests were performed using contact method with 5 MHz in frequency and 10 mm in diameter of transducer. Amplitude of reflected wave from intermediate layer was used as a quantitative parameter. A set of experiment results show that shear fracture load has a linear correlation with amplitude of reflected wave. Besides, amplitude of reflected wave also has relation with nugget diameter. It could be concluded that ultrasonic contact method could be applied in predicting a shear fracture load. Keywords : Non-destructive prediction, Ultrasonic test, Spot welding, Shear strength

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P-91 STEADY STATE AND LOCA ANALYSIS OF KARTINI REACTOR USING RELAP5/SCDAP CODE: THE ROLE OF PASSIVE SYSTEM

Anhar R. Antariksawan1, Puradwi I. Wahyono2 and Taxwim2

1Center for Nuclear Reactor Safety and Technology - BATAN, Puspiptek Area Building 80, Serpong, Tangerang Selatan 15310

2Center for Accelerator Science and Technology - BATAN, Jl. Babarsari, Yogyakarta email: [email protected]

ABSTRACT

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed. Keywords: research reactor, safety, thermal-hydraulic, LOCA, RELAP5

P-92 POWER PEAKING EFFECT OF OTTO FUEL SCHEME PEBBLE BED REACTOR

T. Setiadipura, Suwoto, Zuhair, S. Bakhri, G.R. Sunaryo Center for Nuclear Reactor Technology and Safety – BATAN

Puspiptek Area, Office Building No. 80, Serpong, Tangerang Selatan 15310, Indonesia

ABSTRACT Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effects are stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

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P-93 THE EFFECT OF ZINC INJECTION ON THE INCREASING OF INCONEL 600 TT CORROSION RESISTANCES

Febrianto, Sriyono, Surip Widodo, Geni Rina S

Puspiptek Area Building 80, Serpong, South Tangerang City, 15310, INDONESIA Email: [email protected]

ABSTRACT

Many failures were found in reactor pressure vessel head penetration (RPV) head material. Those failures caused by boric acid corrosion, and from visual examination were found a big hole and white deposit crystal of boric acid during shutdown maintenance at David Besse reactor. Zinc Oxide addition in BWR reactor known as Zinc Injection that has purposed to reduce radiation exposure cause of Hydrogen addition. Beside reducing the radiation exposure, Zinc injection also has an effect in reducing material corrosion. The purpose of study is to determine the effect of zinc addition, boric acid, temperature also the effects of Cobalt Nitrate and Zinc Oxide addition to Inconel 600 TT as RPV head penetration material. The result in the BWR reactor experience will be implementated at PWR reactor, weather zinc oxide addition also has an effect in reducing the corrosion of Inconel 600. The method that used in this research is to observe the corrosion rates for Inconel 600 material using Potentiostat. Examination were conducted in 30 , 40, 60, 70, 80 and 80 °C using 1000, 1500, 2000, 2500 and 3000 ppm boric acid concentration. The results showed that the corrosion rate for the material were very small, but the highest corrosion rate occurred in 3000 ppm boric acid concentration at 90 °C with Cobalt Nitrate addition, around 5.210 x 10-1 mpy. In the same condition at 3000 ppm boric acid concentration for temperature at 90 °C, Inconel 600 TT corrosion rate is smaller with Zinc oxide addition, around 4.631 x 10-1 mpy.

P-94 NEUTRON RADIATION DAMAGE ESTIMATION IN THE CORE STRUCTURE BASE METAL OF RSG GAS

S A Santa and Suwoto Center for Nuclear Reactor Safety and Technology , Puspiptek Complex, Building No. 80, Serpong, Tangerang

Selatan, 15310, Indonesia Email: [email protected]

ABSTRACT

Radiation damage in core structure of the Indonesian RGS GAS multi purpose reactor resulting from the reaction of fast and thermal neutrons with core material structure was investigated for the first time after almost 30 years in operation. The aim is to analyze the degradation level of the critical components of the RSG GAS reactor so that the remaining life of its component can be estimated. Evaluation results of critical components remaining life will be used as data ccompleteness for submission of reactor operating permit extension. Material damage analysis due to neutron radiation is performed for the core structure components made of AlMg3 material and bolts reinforcement of core structure made of SUS304. Material damage evaluation was done on Al and Fe as base metal of AlMg3 and SUS304, respectively. Neutron fluences are evaluated based on the assumption that neutron flux calculations of U3Si8-Al equilibrium core which is operated on power rated of 15 MW. Calculation result using SRAC2006 code of CITATION module shows the maximum total neutron flux and flux >0.1 MeV are 2.537E+14 n/cm2/s and 3.376E+13 n/cm2/s, respectively. It was located at CIP core center close to the fuel element. After operating up to the end of #89 core formation, the total neutron fluence and fluence >0.1 MeV were achieved 9.063E+22 and 1.269E+22 n/cm2, respectively. Those are related to material damage of Al and Fe as much as 17.91 and 10.06 dpa, respectively. Referring to the life time of Al-1100 material irradiated in the neutron field with thermal flux/total flux=1.7 which capable of accepting material damage up to 250 dpa, it was concluded that RSG GAS reactor core structure underwent 7.16% of its operating life span. It means that core structure of RSG GAS reactor is still capable to receive the total neutron fluence of 9.637E+22 n/cm2 or fluence >0.1 MeV of 5.672E+22 n/cm2. Keywords: RSG GAS, critical component, irradiation embrittlement, material damage, ageing component.

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P-95 ANALYSIS RESPONS TO THE IMPLEMENTATION OF NUCLEAR INSTALLATIONS SAFETY CULTURE USING AHP-TOPSIS

J Situmorang, I.Kuntoro, S Santoso, M Subekti, G.R Sunaryo Center for Nuclear Reactor Technology and Safety, BATAN

Puspiptek Complex, Building No.80, Serpong, Tangerang Selatan 15310 Indonesia E-mail: [email protected]

ABSTRACT

An analysis of responses to the implementation of nuclear installations safety culture has been done using AHP (Analitic Hierarchy Process) - TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). Safety culture is considered as collective commitments of the decision-making level, management level, and individual level. Thus each level will provide a subjective perspective as an alternative approach to implementation. Furthermore safety culture is considered by the statement of five characteristics which in more detail form consist of 37 attributes, and therefore can be expressed as multi-attribute state. Those characteristics and or attributes will be a criterion and its value is difficult to determine. Those criteria of course, will determine and strongly influence the implementation of the corresponding safety culture. To determine the pattern and magnitude of the influence is done by using a TOPSIS that is based on decision matrix approach and is composed of alternatives and criteria. The weight of each criterion is determined by AHP technique. The data used are data collected through questionnaires at the workshop on safety and health in 2015. .Reliability test of data gives Cronbah Alpha value of 95.5% which according to the criteria is stated reliable. Validity test using bivariate correlation analysis technique between each attribute give Pearson correlation for all attribute is significant at level 0,01. Using confirmatory factor analysis gives Kaise-Meyer-Olkin of sampling Adequacy (KMO) is 0.719 and it is greater than the acceptance criterion 0.5 as well as the 0.000 significance level much smaller than 0.05 and stated that further analysis could be performed. As a result of the analysis it is found that responses from the level of decision maker (second echelon) dominate the best order preference rank to be the best solution in strengthening the nuclear installation safety culture, except for the first characteristics, safety is a clearly recognized value. The rank of preference order is obtained sequentially according to the level of policy maker, management and individual or staff. Keywords: Safety Culture, AHP, TOPSIS

P-96 ANALYSIS ON OPERATING PARAMETER DESIGN TO STEAM METHANE REFORMING IN HEAT APPLICATION RDE

Sukmanto Dibyo, Geni Rina Sunaryo, Syaiful Bakhri, Zuhair, Ign.Djoko Irianto

PTKRN Batan Gedung 80 kawasan Puspiptek Serpong Email: [email protected]

ABSTRACT

The high temperature reactor has been developed with various power capacities and can produce electricity and heat application. One of heat application is used for hydrogen production. Most hydrogen production occurs by steam reforming that operated at high temperature. This study aims to analyze the feasibility of heat application design of RDE reactor in the steam methane reforming for hydrogen production using the ChemCAD software. The outlet temperature of cogeneration heat exchanger is analyzed to be applied as a feed of steam reformer. Furthermore, the additional heater and calculating amount of fuel usage are described. Results show that at a low mass flow rate of feed, its can produce a temperature up to 480oC. To achieve the temperature of steam methane reforming of 850oC the additional fired heater was required. By the fired heater, an amount of fuel usage is required depending on the Reformer feed temperature produced from the heat exchanger of the cogeneration system. Keywords: RDE, heat application, hydrogen production, steam methane reformer, additional heater.

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P-97 RADIOLOGICAL RISK ESTIMATION OF 137CESIUM ON A NEAR SURFACE DISPOSAL FACILITY BY USING RESRAD ONSITE APPLICATION

Budi Setiawan1*, Suci Prihastuti2, Setyo S Moersidik3

1Center for Radioactive Waste Technology-National Nuclear Energy Agency, PUSPIPTEK Bld 71 3rd Floor, Serpong-Tangerang 15310, Indonesia

2Directorate for Inspection of Nuclear Installation and Material-Nuclear Energy Regulatory Agency, Jl.Gadjahmada No. 8, Jakarta 10120, Indonesia

3Department of Civil Engineering-University of Indonesia, Kampus UI, Depok 16424, Indonesia *Corresponding author email: [email protected]

ABSTRACT

The operational of near surface disposal facility during waste packages loading activity into the facility, or in a monitoring activity around disposal facility is predicted to give a radiological risk to radiation workers. The thickness of disposal facility cover system affected the number of radiological risk of workers. Due to this reason, a radiological risk estimation needs to be considered. RESRAD onsite code is applied for this purpose by analyzing the individual accepted dose and radiological risk data of radiation workers. The obtained results and then are compared with radiation protection reference in accordance with national regulation. In this case, the data from the experimental result of Karawang clay as host of disposal facility such as Kd value of 137Cs was used. Results showed that the thickness of the cover layer of disposal facility affected to the radiological risk which accepted by workers in a near surface disposal facility. Keywords: near surface disposal, 137Cs, cover system, Kd, radiological risk

P-98 THE SIMULATOR DEVELOPMENT FOR RDE REACTOR

Muhammad Subekti, Syaiful Bakhri, Geni Rina Sunaryo Center for Nuclear Reactor Technology and Safety, BATAN

Puspiptek Complex, Building No.80, Serpong, Tangerang Selatan 15310, Indonesia. Email: [email protected]

ABSTRACT BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

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P-99 MODELLING THE RADIOLYSIS OF RSG-GAS PRIMARY COOLING WATER: PRELIMINARY STUDY

Sofia Loren Butarbutar, Rahayu Kusumastuti, M. Subekti, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy of Agency (BATAN), Puspiptek Area Building 80, Banten, Indonesia


ABSTRACT Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water -rays and neutron radiolysis (pH=7) at temperature between 25 and 50 oC. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic -rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

P-100 INFLUENCE OF SOL CONCENTRATION TO PARTICLE DIAMETER OF CERIUM STABILIZED ZIRCONIUM MADE BY EXTERNAL GELATION

Sukarsono, Meniek Rahmawati, Sri Rinanti S, Dedy Husnurrofiq, Kristanti And Ariyani Dewi K

PTBBN BATAN SERPONG TANGERANG Email: [email protected]

ABSTRACT

Cerium Stabilized Zirconium gel has been prepared with external gelation. As the raw materials was used ZrO(NO3)2 and Ce(NO3)4 nitrate salt which is dissolved with water into Zr-Ce nitrate mixture. The concentration of the nitrate salt mixture in the sol solution was varied by varying the concentration of zirconium and cerium nitrate in the sol solution and the addition of PVA and THFA to produce a sol with a viscosity of 40-60 cP. The viscosity range 40-60cP is the viscosity of the sol solution that was easy to produce a good gel in the gelation apparatus. Sol solution was gelated in a gelation column equipped with a 1 mm diameter drip nozzle, nozzle vibrator to adjust the best frequency and amplitudo vibration, a flowmeter to measure the flow rate of sol, flowing of NH3 gas to presolidification process, a column containing gelation medium and gel container to collect gel product. Gel obtained from the gelation process than processed with aging, washing, drying and calcination to get round gel and not broken at calcination up to 500oC. From the gelation processes that has been done, it can be seen that with the presolidification process can be obtained a round gel and without presolidifikasi process, produce not round gel. In the process of ageing to get not broken gel, ageing was done on the rotary flask so that during the ageing, gels rotate in gelation media. Gels, then be washed by dilute ammonium nitrate, demineralized water dan iso prophil alcohol. The washed gel was then dried by vacuum drying to form pores on the gel which is the path for the gases resulting from decomposition of the gel. Vacuum drying can prevent craking because the pores allow the gel to release the decomposition of the material during heating. Larger the concentration of nitric metal in sol solution, yields a gel with a larger diameter and allows us to plan the diameter of the sintered particles to be made.

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P-101 REACTIVITY COEFFICIENT CALCULATION FOR AP1000 REACTOR USING THE NODAL3 CODE

Surian Pinem, Tagor Malem Sembiring, Deswandri, Geni Rina Sunaryo Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesian (BATAN),

Kawasan Puspiptek Building No. 80,Tangerang Selatan 15310, Banten, Indonesia Email: [email protected]

ABSTRACT The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/oC, -1.00518 pcm/oC to 1.00649 pcm/oC and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 oC. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value. Keywords: reactivity coefficient, inherent safety, NODAL3, SRAC2006, AP1000

P-102 STUDY ON CHARACTERISTIC OF TEMPERATURE COEFFICIENT OF REACTIVITY FOR PLUTONIUM CORE OF PEBBLED BED REACTOR

Zuhair, Suwoto, T. Setiadipura, S. Bakhri, G.R. Sunaryo Center for Nuclear Reactor Technology and Safety – National Nuclear Energy Agency of Indonesia (BATAN),

Puspiptek Complex, OB No. 80, Serpong, Tangerang Selatan 15310, Indonesia, Tel. (021)756-0912, Fax. (021)756-0913


ABSTRACT As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the keff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields keff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation. Keywords: temperature coefficient of reactivity, plutonium core, pebble bed reactor, MCNPX, ENDF/B-VII

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P-103 EFFECTS OF THE APPLICATION OF THE NEW NUCLEAR DATA LIBRARY ENDF/B TO THE CRITICALITY ANALYSIS OF AP1000

Iman Kuntoro,T.M. Sembiring, Deswandri, G.R. Sunaryo

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Gd. No. 80 Serpong , Tangerang Selatan 15310 Indonesia


ABSTRACT Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1. Keywords: criticality analysis, AP1000, ENDF/B, MCNP

P-104 A PRELIMINARY DESIGN OF APPLICATION OF WIRELESS IDENTIFICATION AND SENSING PLATFORM ON EXTERNAL BEAM RADIOTHERAPY

Heranudin1,2 and S. Bakhri3 1 Center for Radioisotope and Radiopharmaceutical Technology, National Nuclear Energy Agency of Indonesia,

South Tangerang, Indonesia 2 School of Physics, University of Wollongong, New South Wales, Australia

3 Center for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia, South Tangerang, Indonesia


ABSTRACT A linear accelerator (linac) is widely used as a means of radiotherapy by focusing high-energy photons in the targeted tumor of patient. Incorrectness of the shooting can lead normal tissue surrounding the tumor received unnecessary radiation and become damaged cells. A method is required to minimize the incorrectness that mostly caused by movement of the patient during radiotherapy process. In this paper, the Wireless Identification and Sensing Platform (WISP) architecture was employed to monitor in real time the movement of the patient's body during radiotherapy process. In general, the WISP is a wearable sensors device that can transmit measurement data wirelessly. In this design, the measurement devices consist of an accelerometer, a barometer and an ionizing radiation sensor. If any changes in the body position which resulted in incorrectness of the shooting, the accelerometer and the barometer will trigger a warning to the linac operator. In addition, the radiation sensor in the WISP will detect unwanted radiation and that can endanger the patient. A wireless feature in this device can ease in implementation. Initial analyses have been performed and showed that the WISP is feasible to be applied on external beam radiotherapy. Keywords: Radiotherapy, Radiation dose, Accuracy, Wearable sensor.

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P-105 LEACHING KINETIC OF Nd. Y, Pr AND Sm IN RARE EARTH HYDROXIDE (REOH) USE NITRIC ACID

MV Purwani and Suyanti

Center for Accelerator Science and Technology National Nuclear Energy Agency Email : [email protected]

ABSTRACT

The studied parameters were leaching temperature (60 – 90 C) and leaching time (0-15 minutes). The purpose of this research were to determine the reaction order and activation energy of the reaction of Y(OH)3, Nd(OH)3, Pr(OH)3 and Sm(OH)3 with HNO3. From the resulting data can be concluded that the leaching process were strongly influenced by the time and temperature process. Leaching rare earth hydroxide (REOH) using nitric acid follows second order. At leaching 10 grams of REOH using 40 ml HNO3 0.0576 mol were obtained maximum conversion at 90 C and leaching time 15 minutes for Y was 0.95 (leaching efficiency was 95%), for Nd was 0.97 ( leaching efficiency was 97%), for Pr was 0.94 (leaching efficiency was 94%) and for Sm was 0.94 (leaching efficiency was 94%). The largest activation energy was Y of 23.34 kJ /mol followed by Pr of 20.00 kJ /mol, Sm of 17.94 kJ /mol and the smallest was Nd of 16.39 kJ /mol. The relationship between the rate constant of the reaction with T for Y was kY = 338.26 e -23,34 / RT, for Nd was kNd = 33.,69 e - 16,39 / RT, for Pr was kPr = 102.04 e -20 / RT and for Sm adalah was kSm = 50,16 e –17,94/ RT Keywords: REOH, HNO3, leaching

P-106 PRELIMINARY DEVELOPMENT OF ONLINE MONITORING ACOUSTIC EMISSION

SYSTEM FOR THE INTEGRITY OF RESEARCH REACTOR COMPONENTS

S Bakhri1, E Sumarno1, R Himawan1, T Y Akbar2, M. Subekti1, G. R. Sunaryo1 1Centre for Nuclear Reactor Technology and Safety, Puspiptek Complex, OB 80, Serpong, Setu, Tangerang

Selatan, Indonesia 2Jurusan Fisika, Fakultas Matematika dan Ilmu Pengetahuan Alam, Universitas Sriwijaya


ABSTRACT Three research reactors owned by BATAN have been more than 25 years. Aging of (Structure, System and Component) SSC which is mainly related to mechanical causes become the most important issue for the sustainability and safety operation. Acoustic Emission (AE) is one of the appropriate and recommended methods by the IAEA for inspection as well as at the same time for the monitoring of mechanical SSC related. However, the advantages of AE method in detecting the acoustic emission both for the inspection and the online monitoring require a relatively complex measurement system including hardware software system for the signal detection and analysis purposes. Therefore, aim of this work was to develop an AE system based on an embedded system which capable for doing both the online monitoring and inspection of the research reactor’s integrity structure. An embedded system was selected due to the possibility to install the equipment on the field in extreme environmental condition with capability to store, analyses, and send the required information for further maintenance and operation. The research was done by designing the embedded system based on the Field Programmable Gate Array (FPGA) platform, because of their execution speed and system reconfigurable opportunities. The AE embedded system is then tested to identify the AE source location and AE characteristic under tensile material testing. The developed system successfully acquire the AE elastic waveform and determine the parameter-based analysis such as the amplitude, peak, duration, rise time, counts and the average frequency both for the source location test and the tensile test.

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P-107 SOLVENT SELECTION FOR EXTRACTION OF NEODYMIUM CONCENTRATES OF MONAZITE SAND PROCESSED PRODUCT

Moch Setyadji. MV Purwani

Centre for Accelerator Science and Technology – BATAN Yogyakarta Jalan Babarsari. Kotak Pos 6101 ykbb.Yogyakarta 55281

E-mail : [email protected]

ABSTRACT The extraction of neodymium concentrates of monazite sand processed product has been done. The objective of this investigation was to determine the best solvent to separate Nd from Nd concentrate. As an aqueous phase was Nd(OH)3 concentrated in HNO3 and as solvent or the organic phase was trioctylamine (TOA). tryibuthyl phosphate (TBP). trioctylphosphine oxyde (TOPO)and di-ethyl hexyl phosphoric acid (D2EHPA) in kerosene. The investigated variables were HNO3 concentration. feed concentration. solvent concentration or solvent in kerosene. time and stirring speeds. From the investigation on the selection of solvent for the extraction of Nd(OH)3 concentrate with various solvents. it was concluded that the extraction of Nd could be carried out by using TBP or TOA. Extraction of Nd using TOA at the optimum HNO3 concentration of 2M. feed concentration of 5 gram/10 mL. TOA in kerosene concentration of 6 %. stirring time of 15 minutes. stirring speed of 200 rpm was chosen if the Y concentration in Nd concentrate is small. In these condition DNd obtained was 0.65; extraction efficiency of Nd (ENd)=37.10%. the concentrations of Nd2(C2O4)3 = 67.14%. Ce2(C2O4)3 = 1.79%. La2(C2O4)3 = 1.37% and Y2(C2O4)3 = 24.70%. Extraction of Nd using TBP at the optimum HNO3 concentration of 1M. feed concentration of 5 gram/10 m. the TBPconcentration in kerosene of 15%. stirring time of 15 minutes and stirring speed of 200 rpm was chosen if the Ce concentration in Nd concentrate is small. In these condition DNd obtained was 0.20. extraction efficiency of Nd (ENd)=17%. concentration of Nd2(C2O4)3 =70.84%. Ce2(C2O4)3=15.53%. La2(C2O4)3 =0.00% and Y2(C2O4)3 = 8.63%. Keywords: Solvent selection. neodymium concentrates. tri oxtyl amine (TOA). tributyl phosphate (TBP). diethyl hexyl phosphoric acid (D2EHPA) dan trioxthyl phosphone oxyde (TOPO)

O-1 RELIABILITY ANALYSIS OF RSG-GAS PRIMARY COOLING SYSTEM TO SUPPORT AGING MANAGEMENT PROGRAM

Deswandri, M.Subekti, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety – BATAN Kawasan Puspiptek Gd. 80, Setu, Tangerang Selatan


ABSTRACT Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed. One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

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O-2 SENSOR FAILURE DETECTION OF FASSIP SYSTEM USING PRINCIPAL COMPONENT ANALYSIS

Sudarno, Mulya Juarsa, Kussigit Santosa, Deswandri, Geni Rina Sunaryo

Center for Nuclear Reactor Technology and Safety, Puspiptek Area Building 80, Serpong, South Tangerang City, 15310, INDONESIA


ABSTRACT In the nuclear reactor accident of Fukushima Daiichi in Japan, the damages of core and pressure vessel were caused by the failure of its active cooling system (diesel generator was inundated by tsunami). Thus researches on passive cooling system for Nuclear Power Plant are performed to improve the safety aspects of nuclear reactors. The FASSIP system (Passive System Simulation Facility) is an installation used to study the characteristics of passive cooling systems at nuclear power plants. The accuracy of sensor measurement of FASSIP system is essential, because as the basis for determining the characteristics of a passive cooling system. In this research, a sensor failure detection method for FASSIP system is developed, so the indication of sensor failures can be detected early. The method used is Principal Component Analysis (PCA) to reduce the dimension of the sensor, with the Squarred Prediction Error (SPE) and statistic Hotteling criteria for detecting sensor failure indication. The results shows that PCA method is capable to detect the occurrence of a failure at any sensor. Keywords: Sensor, fault detection, FASSIP, Principal Component Analysis

O-3 PRELIMINARY ANALYSIS OF HIGH-FLUX RSG-GAS TO TRANSMUTE AM-241 OF

PWR’S SPENT FUEL IN ASIAN REGION

M Budi Setiawan and S Kuntjoro Center for Nuclear Reactor Technology and Safety, Puspiptek Complex, building no. 80, Serpong, Tangerang

Selatan 15410 Indonesia Email: [email protected]

ABSTRACT

A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

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O-4 KR-85M ACTIVITY AS BURNUP MEASUREMENT INDICATOR IN A PEBBLE BED

REACTOR BASED ON ORIGEN2.1 COMPUTER SIMULATION

I Husnayani, P M Udiyani, S Bakhri, G R Sunaryo

Center for Nuclear Reactor Technology and Safety, Puspiptek Complex, building no. 80, Serpong, Tangerang Selatan 15314 Indonesia


ABSTRACT Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step, in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

O-5 OPERATOR SUPPORT SYSTEM DESIGN FOR THE OPERATION OF RSG-GAS

RESEARCH REACTOR

S Santoso, J Situmorang , S Bakhri, M Subekti, G.R Sunaryo Center for Nuclear Reactor Technology and Safety-BATAN, Puspiptek Complex, Building no.80,

Serpong,Tangerang Selatan 15310, Indonesia Email: [email protected]

ABSTRACT

The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room. Keywords: operator support system, RSG-GAS, control room

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O-6 V&V PLAN FOR ESF-CCS BASED FPGA USING SYSTEM ENGINEERING APPROACH

Restu Maerani, Joyce Mayaka, Mohamed El Akrat, Jung Jae Cheon KEPCO International Nuclear Graduate School, Ulsan, South Korea


ABSTRACT Instrumentation and Control systems play an important role in maintaining the safety of Nuclear Power Plant (NPP) operation. Therefore, the plan for the replacement of Programmable Logic Controller (PLC) based Engineered Safety Feature – Component Control System (ESF-CCS) with Field Programmable Gate Arrays (FPGA) because of considering the importance to meet the safety criteria for nuclear reactors that should prevent the common-cause failure is needed. By using system engineering approach, which are ensuring traceability in every phase in the lifecycle, from system requirements, design implementation to verification and validation, the system development is guaranteed in line with the regulatory requirements. The Verification process will ensure that the customer and stakeholder's needs are satisfied in a high quality, trustworthy, cost efficient and schedule compliant manner throughout a system's entire life cycle. The benefit of the V&V plan is to ensure that the FPGA based ESF-CCS is correctly built, and to ensure that the measurement of performance has positive feedback that “do we do the right thing” during the re-engineering process of the development of an FPGA based ESF-CCS. Keywords: ESF-CCS, FPGA, Verification, Validation

O-7 PRELIMINARY INVESTIGATION OF TIME REMAINING DISPLAY ON THE COMPUTER-BASED EMERGENCY OPERATING PROCEDURE

T J Suryono and A Gofuku

Graduate School of Natural Science and Technology, Okayama University, 3-1-1 Tsushimanaka, Kita-ku, Okayama, 700-8530, Japan

Email: [email protected] ABSTRACT

One of the important thing in the mitigation of accidents in nuclear power plant accidents is time management. The accidents should be resolved as soon as possible in order to prevent the core melting and the release of radioactive material to the environment. In this case, operators should follow the emergency operating procedure related with the accident, in step by step order and in allowable time. Nowadays, the advanced main control rooms are equipped with computer-based procedures (CBPs) which is make it easier for operators to do their tasks of monitoring and controlling the reactor. However, most of the CBPs do not include the time remaining display feature which informs operators of time available for them to execute procedure steps and warns them if the they reach the time limit. Furthermore, the feature will increase the awareness of operators about their current situation in the procedure. This paper investigates this issue. The simplified of emergency operating procedure (EOP) of steam generator tube rupture (SGTR) accident of PWR plant is applied. In addition, the sequence of actions on each step of the procedure is modelled using multilevel flow modelling (MFM) and influenced propagation rule. The prediction of action time on each step is acquired based on similar case accidents and the Support Vector Regression. The derived time will be processed and then displayed on a CBP user interface. Keywords: computer-based procedure, time remaining, multilevel flow modelling, support vector regression, PWR type reactor, steam generator tube rupture

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O-8 DESIGN OF FISHING BOAT FOR PELABUHANRATU FISHERMEN AS ONE OF EFFORT

TO INCREASE PRODUCTION OF CAPTURE FISHERIES

Iswadi Nur, Purwo Joko Suranto Study Program of Naval Architecture1, Engineering Faculty of UPN ”Veteran” Jakarta


ABSTRACT Design of fishing boat for Pelabuhanratu fisherman as one of effort to increase production of capture fisheries. The fishing boat should be proper for the characteristic of its service area, as ;capacity of fishing boat up to 60 GT, the fishing boat has minimum 6 fish holds and location of fish hold in the middle body, the fishing boat hull has the bilge keel plate, and the material of hull fishing boat to be made of wooden, steel, aluminium, or fiberglass. Main dimesion of fishing boat is Length Over All = 25,436 m, Breadth = 4,55 m, Draft = 1,6 m, Speed = 12,5 knots. The research had been known every thing that will be supporting the production of capture fisheries like ; amount of fish production = 25,030 ton per day, the fishing port capacity approximately 268,957GT per day, the area of fishing port < 30 hectares, the zone of fish processing industry had not completed, therefore all data research result less than standard of Oceanic Fising Port. So Pelabuhanratu National Fishing Port can not be changed to Oceanic Fishing Port. Keywords : fishing boat, national fishing port, oceanic fishing port

O-9 OPTIMIZATION OF PARAMETERS FOR MANUFACTURE NANOPOWDER BIOCERAMICS AT MACHINE PULVERISETTE 6 BY TAGUCHI AND ANOVA METHOD

Hendri Van Hotena, Gunawarmanb, Ismet Hari Mulyadib, Afdhal Kurniawan

Mainila dan Putra Bismantoloa a)Mechanical Engineering, Bengkulu University, Indonesia b)Mechanical Engineering, Andalas University, Indonesia

Corresponding author’s: [email protected]

ABSTRACT This research is about manufacture nanopowder Bioceramics from local materials used Ball Milling for biomedical applications. Source materials for the manufacture of medicines are plants, animal tissues, microbial structures and engineering biomaterial. The form of raw material medicines is a powder before mixed. In the case of medicines, research is to find sources of biomedical materials that will be in the nanoscale powders can be used as raw material for medicine. One of the biomedical materials that can be used as raw material for medicine is of the type of bioceramics is chicken egg shells. This research will develop methods for manufacture nanopowder material from chicken egg shells with Ball Milling using the Taguchi method and ANOVA. Eggshell milled using a variation of Milling rate on 150, 200 and 250 rpm, the time variation of 1, 2 and 3 hours and variations the grinding balls to egg shell powder weight ratio 1: 6, 1: 8, 1: 10. Before milled with Ball Milling crushed eggshells in advance and calcinate to a temperature of 900oC. After the milled material characterization of the fine powder of eggshell using SEM to see its size. The result of this research from analyzing the parameter contribution process to size measure of chicken's eggshell powder. Milling speed, milling time and ball to powder weight of ratio have contribution successively equal to 60.82%, 30.76% and 6.64% by error equal to 1.78%.

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O-10 STUDY ON GOLD AND BASE METAL OCCURRENCE IN ULUWAI PROSPECT, WESTERN LATIMOJONG MOUNTAIN, SOUTH SULAWESI

Adi Maulana1, Asri Jaya1, Akira Imai2

1Department of Geology, Faculty of Engineering, Hasanuddin University 2Department of Earth Resources Engineering, Akita University


ABSTRACT Uluwai Prospect is located in the northern part of South Arm of Sulawesi along the eastern part of the Kalosi Fold Belt and Latimojong Mountain. The area is generally characterized by moderate to rugged topography area with elevation in the range of 700 to 1400 m above sea level in the mountainous complex called Latimojong Mountain Complex. The mineralization is characterized by a relatively simple sulphide ore mineral assemblage consists of pyrite, sphalerite and chalcopyrite. Samples were collected in areas showing abundant sulphide minerals where younger faults cut the bedding and foliation of country rocks. A number of silicified zones have been observed, as well as float material containing disseminated pyrite, chalcopyrite, and sphalerite with hematite, goethite and limonite. Some alteration types have been observed including sericitization, albitization, carbonatization and silisification. The samples collected indicated that the mineralisation is contained within metasedimentary (sandstone to mudstone) and greenschist. Geochemical analyses from 16 samples including 5 stream sediment samples indicated that the most promising mineralization occur in the prospect area are copper (Cu) and zinc (Zn). This is also supported by the abundance of chalcopyrite and sphalerite in some highly altered samples. Assaying of the collected samples revealed most of samples contain relatively low gold (Au) concentration. However, two samples contain 0.007 and 0.01 ppm of Au. In the mineralized area, Zn concentrations are up to 134 ppm, Cu up to 120 ppm and Pb up to 18 ppm and As up to 70 ppm. There is no clear relationship that exists between Au and the base metals except that one of the samples with highest Au values tend to have high Zn and As. This unclear pattern also shown by Cu, Pb and Zn. Base metal concentration in stream sediment samples show a relatively stable pattern than in rock samples. Arsenic tends to be elevated in base metal rich samples. Sb and Mo are relatively low in all sample type. However, Mo values will be high in the samples which contain highest Cu and Zn. Keywords: Gold, Base metal, Uluwai, Latimojong, South Sulawesi

O-11 A SUB-TARGET APPROACH TO THE KINODYNAMIC MOTION CONTROL OF A WHEELED MOBILE ROBOT

Kimiko Motonaka*, Keigo Watanabe**, and Shoichi Maeyama**

*Kansai University **Okayama University


ABSTRACT A mobile robot with two independently driven wheels is popular, but it is difficult to stabilize it by a continuous controller with a constant gain, due to its nonholonomic property. It is guaranteed that a nonholonomic controlled object can always be converged to an arbitrary point using a switching control method or a quasi-continuous control method based on an invariant manifold in a chained form. From this, the authors already proposed a kinodynamic controller to converge the states of such a two-wheeled mobile robot to the arbitrary target position while avoiding obstacles, by combining the control based on the invariant manifold and the harmonic potential field (HPF). On the other hand, it was confirmed in the previous research that there is a case that the robot cannot avoid the obstacle because there is no enough space to converge the current state to the target state. In this paper, we propose a method that divides the final target position into some sub-target positions and moves the robot step by step, and it is confirmed by the simulation that the robot can converge to the target position while avoiding obstacles using the proposed method.

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O-12 KINEMATICS ANALYSIS OF END EFFECTOR FOR CARRIER ROBOT OF FEEDING BROILER CHICKEN SYSTEM

Rafiuddin Syam1, Hairul Arsyad1, Ruslan Bauna1, Ilyas Renreng1, Syaeful Bakhri2 1Mechanical Engineering Department, Faculty of Engineering, Hasanuddin University,

Indonesia 2Center for Nuclear Reactor Technology and Safety – National Nuclear Energy

Agency- BATAN, Indonesia email: [email protected]

ABSTRACT

The demand for commodities, especially Broiler chicken farms are increasing, the volume of feed requirements Broiler chickens increased with age up to the age of 30-57 days required feed 3,829 grams /day/head, so if the chicken population is 3,000 needed transporting feed 11 487 kg/day, This research aims to produce a robot capable of transporting feed in the top of the cage by using a control system so as to make efficient use of manpower. Design robot performed using software design three-dimensional Solidworks2010, process of making the robot is started with the design manufacture three (3) units of mechanical systems (mechanical system for holder feed, mechanical systems for lifter feed and mechanical systems for transporting feed), then do the design process framework as a component buffer so that the mechanical system will work properly and safely when the robot operates. Furthermore, the manufacture of electronic circuits and control are using Arduino Mega microcontroller. After assembling all components mechanical systems and installation of electronic systems and control, then experiments to evaluate the performance of the robot have been made. The results of experiments showed that all components work well according to plan, in particular the speed and acceleration of end effector motion so it can hold and release the feed well. This strongly supports the robots perform tasks in accordance with the intent, i.e., holding, lifting and moving feed. Keywords: End effector, Mechanical System for Broilers System Arduino, control design

O-13 THE ULTIMATE STRENGTH OF DOUBLE HULL OIL TANKER DUE TO GROUNDING AND COLLISION

Samuel Izaak Latumahina, Ganding Sitepu, Muhammad Zubair Muis Alie

Department of Naval Engineering, Faculty of Engineering, Hasanuddin University Makassar, South Sulawesi, Indonesia

Email: [email protected] ABSTRACT

The damaged tanker by grounding and collision may totally collapse if loss its buoyancy, stability and suffer structural failure. The objective of the present study is to investigate the ultimate strength of double hull oil tanker under vertical bending moments due to grounding and collision. The damages are modelled by removing the elements consist of stiffened and unstiffened plates from the damages part. One-frame space of the double hull oil tanker is taken to be analysed. Two damages cases are considered in the analyses those are grounding and collision. The transversal damage extent for grounding are 10%, 25%, 40% and 55%. The groundings are placed at symmetric position on the outer bottom part. For the case of collision, the vertical damage extent are taken as 10%, 20%, 40% and 60%. The transversal damages extent is taken to be B/16 and it is constant for all collision damages. The investigation of the ultimate strength is performed by the Non-linear Finite Element Analysis method under moment control. The boundary condition is applied with fully constrained on all nodes at the aft-end, while the rigid linked on all nodes is attached at the fore-end with respect to the reference point on the neutral axis. The initial imperfection, welding residual stress and crack are not considered in the analyses. The results obtained by Non-linear Finite Element Analyses for the ultimate strength are compared with the in-house program using Smith’s method implemented in HULLST. The stress distribution and deformation for every case of damages including intact are also discussed in the present study.

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O-14 DEVELOPMENT OF AUTONOMOUS CONTROLLER SYSTEM OF HIGH SPEED UAV FROM SIMULATION TO READY TO FLY CONDITION

Herma Yudhi Irwanto

Indonesian National Institute of Aeronautics and Space Email: [email protected]

ABSTRACT

The development of autonomous controller system that is specially used in our high speed UAV, it’s call RKX-200EDF/TJ controlled vehicle needs to be continued as a step to mastery and to developt control system of LAPAN’s satellite launching rocket. The weakness of the existing control system in this high speed UAV needs to be repaired and replaced using the autonomous controller system. Conversion steps for ready-to-fly system involved controlling X tail fin, adjusting auto take off procedure by adding X axis sensor, procedure of way points reading and process of measuring distance and heading to the nearest way point, developing user-friendly ground station, and adding tools for safety landing. The development of this autonomous controller system also covered a real flying test in Pandanwangi, Lumajang in November 2016. Unfortunately, the flying test was not successful because the booster rocket was blown right after burning. However, the system could record the event and demonstrated that the controller system had worked according to plan.

O-15 A METHOD FOR CALCULATING THE AMOUNT OF MOVEMENTS TO ESTIMATE

THE SELF-POSITION OF MANTA ROBOTS

Takuya Imahama, Keigo Watanabe, Kota Mikuriya and Isaku Nagai Graduate School of Natural Science and Technology

Okayama University, Okayama, Japan Email: [email protected]

ABSTRACT

In recent years, the demand of underwater investigation is increasing in the circumference of a dam, the environmental research of the shallow where approach by ship is difficult, etc. It is known, however, that for man, all over the sea, danger exists mostly, and prolonged diving has a bad influence to a human body. Then, the development of underwater exploration robots that investigate underwater instead of humans is expected. Among underwater exploration robots, it is known that robots imitating aquatic organisms have little influence on underwater environment. Therefore, at this laboratory, a Manta robot using propulsive mechanisms with pectoral fins was developed, imitating the pectoral fin of Manta. Although underwater environmental research needs a function for estimating the self-position, it is not mounted in this Manta robot. This paper explains the amount estimation of movements using optical flows. Especially, a gimbal mechanism is introduced to reduce the influence on the optical flow calculation by pitch motion of the Manta robot. Several experiments are conducted to demonstrate the usefulness of the proposed method.

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O-16 DESIGN OF OMNI DIRECTIONAL REMOTELY OPERATED VEHICLE (ROV)

Rahimuddin1, Hasnawiya Hasan2, Haryanti A Rivai1, Yanu Iskandar3, Claudio P3

1Ph.D, MT, Marine Engineering Department, Hasanuddin University 2M.Eng.Sc, Marine Engineering Department, Hasanuddin University

3Bachelor student, Marine Engineering Department, Hasanuddin University E-mail: [email protected] , [email protected]

ABSTRACT

Nowadays, underwater activities are increased with the increase of oil resources finding. The gap between demand and supply of oil and gas cause engineers to find oil and gas resources in deep water. In other side, high risk of working in deep underwater environment can cause a dangerous situation for human. Therefore, many research activities are developing an underwater vehicle to replace the human’s work such as ROV or Remotely Operated Vehicles. The vehicle operated using tether to transport the signals and electric power from the surface vehicle. Arrangements of weight, buoyancy, and the propeller placements are significant aspect in designing the vehicle’s performance. This paper presents design concept of ROV for survey and observation the underwater objects with interaction vectored propellers used for vehicle's motions.

O-17 MAP GENERATION IN UNKNOWN ENVIRONMENTS BY AUKF-SLAM USING LINE SEGMENT-TYPE AND POINT-TYPE LANDMARKS

Sho Nishihta, Shoichi Maeyama, and Keigo Watanebe

Graduate School of Natural Science and Technology, Okayama University, 3-1-1 Tsushima-naka, Kita-ku, Okayama 700-8350 Japan

Email: [email protected], [email protected], [email protected]

ABSTRACT

Recently, autonomous mobile robots that collect information at disaster sites are being developed. Since it is difficult to obtain maps in advance in disaster sites, the robots being capable of autonomous movement under unknown environments are required. For this objective, the robots have to build maps, as well as the estimation of self-location. This is called a SLAM problem. In particular, AUKF-SLAM which uses corners in the environment as point-type landmarks has been developed as a solution method so far. However, when the robots move in an environment like a corridor consisting of few point-type features, the accuracy of self-location estimated by the landmark is decreased and it causes some distortions in the map. In this research, we propose AUKF-SLAM which uses walls in the environment as a line segment-type landmark. We demonstrate that the robot can generate maps in unknown environment by AUKF-SLAM, using line segment-type and point-type landmarks.

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O-18 SUSTAINABLE MANUFACTURING BY CALCULATING THE ENERGY DEMAND DURING TURNING OF AISI 1045 STEEL

Rusdi Nur, Baso Nasrullah and Asmeati

Department of Mechanical Engineering, Politeknik Negeri Ujung Pandang, Jl. Perintis Kemerdekaan Km. 10 Tamalanrea Makassar, 90245 Indonesia


ABSTRACT Sustainable development will become importance issues for many fields, including production, industry, and manufacturing. In order to achieve sustainable development, industry should be able to perform of sustainable production processes and environmentally friendly. Therefore, there is need to minimize the energy demand in the machining process. This paper presents a calculation method of energy consumption in the machining process, especially turning process which calculated by summing the number of energy consumption, such as the electric energy consumed during the machining preparation, the electrical energy during the cutting processes, and the electrical energy to produce a cutting tool. A case study was performed on dry turning of mild carbon steel using coated carbide. This approach can be used to determine the total amount of electrical energy consumed in the specific machining process. It concluded that the energy consumption will be an increase for using the high cutting speed as well as for the feed rate was increased.

O-19 THE STUDY OF PRODUCTION PERFORMANCE OF WATER HEATER MANUFACTURING BY USING SIMULATION METHOD

M Iqbal1, OAA Bamatraf2 and M Tadjuddin3 1,3Department of Mechanical Engineering, Faculty of Engneering, Syiah Kuala University, Banda Aceh,

Indonesia 2Department of Manufacturing and Material Engineering, Kulliyah of Engineering, International Islamic

University Malaysia Contact email: [email protected]

ABSTRACT

In industrial companies, as demand increases, decision-making to increase production becomes difficult due to the complexity of the model systems. Companies are trying to find the optimum methods to tackle such problems so that resources are utilized and production is increased. One line system of a manufacturing company in Malaysia was considered in this research. The Company produces several types of water heater and each type went into many processes, which was divided into twenty six sections. Each section has several operations. The main type of the product was 10G water heater which is produced most compare to other types, hence it was taken under consideration to be studied in this research. It was difficult to find the critical section that could improve the productions of the company. This research paper employedDelmia Quest software, Distribution Analyser software and Design of Experiment (DOE software) to simulate one model system taken from the company to be studied and to find the critical section that will improve the production system. As a result, assembly of inner and outer tank section were found to be the bottleneck section. Adding one section to the bottleneck increases the production rate by four products a day. The buffer size is determined by the experiment was six items.

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O-20 DESIGN MULTI-SIDES SYSTEM UNMANNED SURFACE VEHICLE (USV) ROCKET

Rafiuddin Syam, Onny Sutresman, Abdullah Mappaita, Ilyas Renreng, Amiruddin, Ardi Wiranata

Mechanical Engineering Department, Engineering Faculty, Hasanuddin University, Indonesia


ABSTRACT. This study aims to design and test USV multislide forms. This system is excellent for maneuvering on the x-y-z coordinates. The disadvantage of a single side USV is that it is very difficult to maneuver to achieve very dynamic targets. While for multi sides system easily maneuvered though x-y-z coordinates. In addition to security defense purposes, multi-side system is also good for maritime intelligence, surveillance. In this case, electric deducted fan with Multi-Side system so that the vehicle can still operate even in reverse condition. Multipleside USV experiments have done with good results. In a USV study designed to use two propulsions.

O-21 THE EFFECTS OF SHIELDED METAL ARC WELDING (SMAW) WELDING ON THE MECHANICAL CHARACTERISTICS WITH HEATING TREATMENT IN S45C STEEL

Munawar1, Hammada Abbas2, Ahmad Yusran Aminy2 1Mechanical Engineering Department Engineering Faculty University of Perjuangan

Republik Indonesia 2Mechanical Engineering Department Engineering Faculty Hasanuddin University


ABSTRACT Steel material has been used mainly for making tooling, automotive components, other household needs, power generators to frame buildings and bridges. This study aimed (1) to analyze the mechanical Characteristics of S45C steel with and without heating treatments, and (2) to analyze the temperature of heating treatment which could result in the maximal strength of S45C steel after the welding process. The research was conducted in the laboratory of mechanical engineering study program, Departement of mechanical Engineering, Christian university of indonesia paulus, makassar. The method used materials, instruments, and the dimensions determination of specimen based on the proposed testing standard, Next, was to determine the mechanical caracteristics of the S45C steel wich had been welded and heated.The tensile specimens, the hardness specimen, the impact specimen, and microstructures of which,each of the 3 specimens was the specimens was the specimen without treatment, the spesimen with the welding wthout heating, and the specimen of 1500, 2500, 3000C. The research results indicated that the treatment process of 1500C, 2500C and 3000C produced the changes of mechanic charateristics with the tensile strength of 42 kgf/mm2 when the temperature had reached 3000C, but at the temperature 3000C, the its toughness would decrease to Hi = 0.836 j/m2 and its hardness would increase to 40.83 at the temperature of 3000C. The value of the maximum strengs was reached at the heating temperature of 3000C for the tensile strength and the hardness, while at the temperature of 3000C its impact value would decrease. Keywords: Steel S45C, SMAW, Heating treatment

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O-22 FLOW RATE AND TEMPERATURE CHARACTERISTICS IN STEADY STATE CONDITION ON FASSIP-01 LOOP DURING COMMISSIONING

M Juarsa, Giarno, A. N. Rohman, G.B. Heru K., J.P. Witoko, D.T. Sony Tjahyani

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan PUSPIPTEK Tangerang Selatan 15340 Banten, Indonesia


ABSTRACT The need for large-scale experimental facilities to investigate the phenomenon of natural circulation flow rate becomes a necessity in the development of nuclear reactor safety management. The FASSIP-01 loop has been built to determine the natural circulation flow rate performance in the large-scale media and aimed to reduce errors in the results for its application in the design of new generation reactors. The commissioning needs to be done to define the capability of the FASSIP-01 loop and to prescribe the experiment limitations. On this commissioning, two scenarios experimental method has been used. The first scenario is a static condition test which was conducted to verify measurement system response during 24 hours without electrical load in heater and cooler, there is water and no water inside the rectangular loop. Second scenario is a dynamics condition that aims to understand the flow rate, a dynamic test was conducted using heater power of 5627 watts and coolant flow rate in the HSS loop of 9.35 LPM. The result of this test shows that the temperature characterization on static test provide a recommendation, that the experiments should be done at night because has a better environmental temperature stability compared to afternoon, with stable temperature around 1C - 3C. While on the dynamic test, the water temperature difference between the inlet-outlets in the heater area is quite large, about 7 times the temperature difference in the cooler area. The magnitude of the natural circulation flow rate calculated is much larger at about 300 times compared to the measured flow rate with different flow rate profiles. Keywords: natural circulation, flow rate, temperature, commissioning, large-scale, FASSIP-01

O-23 THE PNEUMATIC ACTUATORS AS VERTICAL DYNAMIC LOAD SIMULATORS ON MEDIUM WEIGHTED WHEEL SUSPENSION MECHANISM

Simon Ka’ka1, Syukri Himran2, Ilyas Renreng2 and Onny Sutresman2

1Mechanical Engineering Department, State Polytechnic of Ujung Pandang, 90245, South Sulawesi, Indonesia

2Mechanical Engineering Department, Engineering Faculty, Hasanuddin University, 90245, South Sulawesi, Indonesia

The corresponding e-mail address: [email protected]

ABSTRACT Almost all of road damage can be caused by dynamic loads of vehicles that fluctuate according to the type of vehicle that passes through. This study aims to calculate the vertical dynamic load of the vehicle actually occurs on road construction by the mechanism of vehicle wheel suspension. Pneumatic cylinders driven by pressurized air directly load the spring and shock absorber installed on the wheels of the vehicle. The load fluctuations of the medium weight categorized vehicles are determined by the regulation of the amount of pressurized air that enters into the pneumatic cylinder chamber, pushing the piston and connecting rods. The displacement that occurs during compression on the spring and shock absorber, is substituted into the equation of vehicle dynamic load while taking into account the spring stiffness constant, and the fluid or damper gas coefficient. The results show that the magnitude of the displacement when the compression force works has significant influences to the amount of vertical dynamic load of the vehicle that overlies the road construction. The presence of dynamic load of vehicles that fluctuates and repeats, also affects on the reduction of road ability to receive the load. Experimental results using pneumatic actuators instead of real dynamic vehicle loads illustrate the characteristics of the relationship between work pressure and dynamic load. If the working pressure of P2 (bar) is greater, the vertical dynamic load Ft (N) that overloads the road structure is also greater. The associate graphs show that the shock absorber has a greater ability to reduce dynamic load vertically that burden the road structure when compared with the ability of screw spring. Keywords: Pneumatic Cylinder, Dynamic load, Pressurized air, Suspension, Road

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O-24 GEOLOGICAL STUDY AND REGIONAL DEVELOPMENT OF MAMBERAMO RAYA DISCTRICT OF PAPUA PROVINCE, INDONESIA

Adi Tonggiroh, Asri Jaya HS, Ulva Ria Irfan

Geological Engineering Department, Hasanuddin University, Jl. Poros Malino KM.6, Gowa 92171, South Sulawesi, Indonesia Email address: [email protected]

ABSTRACT

The goverment of Mamberamo Raya district was established through Act No. 19 of 2007 dated 15 March 2007 as part of the administrative area of Papua Province. The administrative age of this district is relatively young requires hard work of all components in facing development challenges so that necessary strategic steps of vision and mission of regional development to achieve ideal conditions of spatial which as direction of the desired embodiment in the future. Regional development covers all technical aspects including the geological aspect that the area is located on the morphology of the mountains and Mamberamo watershed. Strategic steps require policy as an action to achieve the goal with the elaboration of operational steps to realize the welfare of peoples equally and sustainably according to the potential physiogeography of Mamberamo watershed. The geological aspect as the consideration of technical that this region belongs to the regional tectonic which is divided into the difference of fault in the north there is Yapen fault and in the south is Mamberamo-Gauttier Fault and also a consideration on the stratigraphic structure of various rock types including the dominance of sedimentary rocks. This study examines geological aspects as an element of earth science in spatial planning in Mamberamo district, especially Kasonaweja and Burmeso. The analysis is presented based on field data, in the form of geographical map data of geological structure, geological map, and earthquake data described by cluster pattern indicating regional motion relationship and rock characteristics that make up Mamberamo watershed. It finds land characteristics controlled by geological structures, rock arrangements and landforms in response to landslide, flood and seismic changes. Keywords: geological aspect,Gauttier Faut, Kasonaweja,Burmeso,Mamberamo

O-25 ANALYZE EXPERIMENT FOR VIGAS AND PERTAMAX TO PERFORMANCE AND EXHAUST GAS EMISSION FOR GASOLINE MOTOR 2000CC

Muhamad As’adi; Diachirta Chrisna Ayu Dwiharpini Tupan

Universitas Pembangunan Nasional “Veteran” Jakarta Jl RS Fatmawati no 1 Pondok Labu Jakarta Selatan Email: [email protected] ; [email protected]

ABSTRACT

In Governor Regulation (APBN – “Anggaran Pendapatan Belanja Negara”) 2015 if Governor will reduce Fuel Oil Subsidy from 48 million Kilo litres to 46 million kilo litres, that is will affect to regulation using BBM for fulfill transportation and industrial needed. One of fuel which is still have enough rest dan their used only for specific used for home needed and food and beverages industry is Liquid Petrolium Gas (LPG). LPG, we called that have differential with the commercial brand is Elpiji and Liquid Gas for Vehicle (LGV or people known with Vigas). LGV have index RON 98 same with the Pertamax and easy to save in portable tank because have low pressure between 8 –15 bar, and pressure for Compressed Natural Gas (CNG) or people known in Indonesian name is Bahan Bakar Gas (BBG) around 200 bar.Develop using LGV in indonesia is slowly because the infrastructure and knowledge people about using LGV for trasnportation industry still less. The purpose and target for this analyze experiment is we get the constanta performance from gasoline motor which used LGV for fuel and Pertamax, so can give knowledge to community if LGV can be use LGV for fuel in transportation industry and more economic. We used experiment method with engine gasoline motor with 2000 cc which is LGV and Pertamax for fuel. The experiment with static experiment tes above Dyno Test. The result is engine perform subscribe Torque, power, fuel consumption. Beside the static test we did the Exhaust Steam Emission. The result is the used LGV with the commercial brand Vigas can increase the maksimum Engine Power 20,86% and Average Power 14,1%, the maximum torque for Motor which is use LGV as fuel is smaller than Motor with Pertamax, the decrease is 0,94%.Using Vigas in Motor can increase the milleage until 6,9% compare with the

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Motor with pertamax.Air Fuel Ratio (AFR) for both of the fuel still below the standard, so still happen waste of fuel, specially in low compression.Using Vigas can reduce the Exhaust Steam Emission especially CO2 Keywords: Gasoline motor ; Vigas ; Pertamax; Engine Perform; Emission.

O-26 COMPUTATIONAL SIMULATION ON FACIAL EXPRESSIONS AND EXPERIMENTAL TENSILE STRENGTH FOR SILICONE RUBBER AS ARTIFICIAL SKIN

Andi Amijoyo Mochtar

Hasanuddin University, Departement of Mechanical Engineering, Jl Perintis Kemerdekaan KM10, Makassar, Indonesia


ABSTRACT The purpose of this research is to make computation on facial expressions and conduct the tensile strength for silicone rubber as artificial skin. Facial expressions were calculated by determining dimension, material properties, number of node elements, boundary condition, force condition, and analysis type. The tensile strength is conducting due to check the proportional force of artificial skin that can be applied on the future of robot facial expression. Combining of calculated and experimental results can generate reliable and sustainable robot facial expression that using silicone rubber as artificial skin.

O-27 RANCANG BANGUN OMNIWHEEL ROBOT SEBAGAI SASARAN TEMBAK DINAMIS

Kamaruddin1, Rafiuddin Syam2

1 Teknik Mesin, Jurusan Teknik Mesin Politeknik Negeri Fakfak, Papua Barat, 98612 2Teknik Mesin, Fakultas Teknik Universitas Hasanuddin Makassar, Makassar, 90245


ABSTRAK RANCANG BANGUN ROBOT PENGGERAK SASARAN TEMBAK. Perkembangan teknologi Automasi dan Mekatronika saat ini yang kian pesat menuntut manusia harus berpacu dengan waktu dimana dibutuhkan suatu alat yang dapat bekerja dengan efektif dan efisien sehingga memudahkan manusia dalam melakukan aktifitasnya. Untuk membuat Robot Penggerak Sasaran Tembak alternatif yang efektif dan efisien serta bekerja secara otomatis, sehingga sasaran yang akan ditembak dapat digerakkan kembali apabila terjadi proses pergantian sasaran dengan sistem yang bekerja secara otomatis sesuai keinginan. Prototipe dari omni wheels robot yang kami buat terdiri dari 4 buah roda dengan bentuk segi empat sama sisi dengan sudut masing-masing roda ke roda lainnya adalah sebesar 900 dari titik tengah. Kata kunci: sasaran tembak, omniwheel, mikrokontroler, kinematika dan dinamika

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O-28 POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER

Ruslan Bauna, Rafiuddin Syam, Hairul Arsyad, Amiruddin DepartemenTeknik Mesin, Fakultas Teknik Mesin, Universitas Hasanuddin

(email: [email protected])

ABSTRAK POLIGON KECEPATAN DAN POLIGON PERCEPATAN END EFFECTOR PADA RANCANG BANGUN ROBOT PENGANGKUT PAKAN AYAM BROILER.Untuk budidaya ayam broiler dengan populasi minimal 3.000 ekor pada umur 30-57 hari membutuhkan usaha pengangkutan pakan di atas kandang minimal 11487 kg. Perencanaan ini bertujuan menghasilkan pemegang pakan (end effector) robot yang mampu memegang pakan dalam karung dengan baik dan efektif serta meletakkan pada tempat yang telah ditentukan secara mandiri dengan menggunakan sistem kontrol. Proses desain pemegang pakanrobot diawali dengan pengambilan data dimensi pakan dalam karung yang akan dipegang oleh end effector robot dan diperoleh dimensi karung pakan 25 cm x 55 cm x 65 cm dengan massa 50 kg, selanjutnya dilakukan proses desain konstruksi end effector yang cocok untuk memegang pakan dengan baik menggunakan program desain tiga dimensi SolidWorks2010. Proses berikutnyaadalah mendesain sistem mekanik pemegang pakan agar gerakan yang dihasilkan sesuai dengan yang diharapkan sehingga tidak merusak wadah pakan dan tetap tercengkram dengan baik saat proses pengangkutan pakan berlangsung. Kemudian, pembuatan rangkaian elektronik dan sistemkendalimenggunakan mikrokontroler Arduino Mega. Setelah dilakukan perakitan semua kompenen sistem mekanik dan pemasangan sistem elektronik dan kontrol, kemudian dilakukan pengujian untuk mengevaluasi kinerja pemegang pakan robot yang telah dibuat. Dari hasil pengujian memperlihatkan bahwa semua komponen bekerja dengan baik sesuai perencanaan, khususnya kecepatan dan percepatan gerak pemegang pakan sehingga mampu memegang dan meletakkan pakan dengan baik dan efektif tanpa merusak wadah pakan. Dengandemikian robot manipulator dapat mengangkat dan membawa pakan secara secara efektif. Kata kunci: mekanisme, pemindah, arduino, kontrol, desain, end effector.

O-29 POWER PEAKING EFFECT OF OTTO FUEL SCHEME PEBBLE BED REACTOR

T. Setiadipura, Suwoto, Zuhair, S. Bakhri, G.R. Sunaryo Center for Nuclear Reactor Technology and Safety – BATAN

Puspiptek Area, Office Building No. 80, Serpong, Tangerang Selatan 15310, Indonesia

ABSTRACT Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effects are stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

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O-30 THE SIMULATOR DEVELOPMENT FOR RDE REACTOR

Muhammad Subekti, Syaiful Bakhri, Geni Rina Sunaryo Center for Nuclear Reactor Technology and Safety, BATAN

Puspiptek Complex, Building No.80, Serpong, Tangerang Selatan 15310, Indonesia. Email: [email protected]

ABSTRACT BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

O-31 MASTER LOGIC DIAGRAM: AN APPROACH TO IDENTIFY INITIATING EVENTS OF HTGRS

J H Purba

Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek, Serpong, Tangerang Selatan – Banten, Indonesia


ABSTRACT Initiating events of a nuclear power plant being evaluated need to be firstly identified prior to applying probabilistic safety assessment on that plant. Various types of master logic diagrams (MLDs) have been proposed for searching initiating events of the next generation of nuclear power plants, which have limited data and operating experiences. Those MLDs are different in the number of steps or levels and different in the basis for developing them. This study proposed another type of MLD approach to find high temperature gas cooled reactor (HTGR) initiating events. It consists of five functional steps starting from the top event representing the final objective of the safety functions to the basic event representing the goal of the MLD development, which is an initiating event. The application of the proposed approach to search for two HTGR initiating events, i.e. power turbine generator trip and loss of offsite power, is provided. The results confirmed that the proposed MLD is feasible for finding HTGR initiating events. Keywords: Master logic diagram, initiating events, high temperature gas cooled reactor, probabilistic safety assessment.

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O-32 PRELIMINARY DEVELOPMENT OF ONLINE MONITORING ACOUSTIC EMISSION SYSTEM FOR THE INTEGRITY OF RESEARCH REACTOR COMPONENTS

S Bakhri1, E Sumarno1, R Himawan1, T Y Akbar2, M. Subekti1, G. R. Sunaryo1

1Centre for Nuclear Reactor Technology and Safety, Puspiptek Complex, OB 80, Serpong, Setu, Tangerang Selatan, Indonesia

2Jurusan Fisika, Fakultas Matematika dan Ilmu Pengetahuan Alam, Universitas Sriwijaya Email: [email protected]

ABSTRACT Three research reactors owned by BATAN have been more than 25 years. Aging of (Structure, System and Component) SSC which is mainly related to mechanical causes become the most important issue for the sustainability and safety operation. Acoustic Emission (AE) is one of the appropriate and recommended methods by the IAEA for inspection as well as at the same time for the monitoring of mechanical SSC related. However, the advantages of AE method in detecting the acoustic emission both for the inspection and the online monitoring require a relatively complex measurement system including hardware software system for the signal detection and analysis purposes. Therefore, aim of this work was to develop an AE system based on an embedded system which capable for doing both the online monitoring and inspection of the research reactor’s integrity structure. An embedded system was selected due to the possibility to install the equipment on the field in extreme environmental condition with capability to store, analyses, and send the required information for further maintenance and operation. The research was done by designing the embedded system based on the Field Programmable Gate Array (FPGA) platform, because of their execution speed and system reconfigurable opportunities. The AE embedded system is then tested to identify the AE source location and AE characteristic under tensile material testing. The developed system successfully acquire the AE elastic waveform and determine the parameter-based analysis such as the amplitude, peak, duration, rise time, counts and the average frequency both for the source location test and the tensile test.

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