PROGRAM BOOK http://icenes2019.fmipa-itb.org

tional Con rna fer te en n c I e th o 9 n 1 October 6-9th, 2019 ICENES2019

E Holiday Inn Resort s m Bali, Indonesia m e e rg t i ys ng S N gy uclear Ener

BAHCESEHIR UNIVERSITY ICENES 2019 Program Book

PREFACE

Dear Colleagues and Participants,

On behalf of the organizing committee, It is my pleasure and honor to welcome all of you: the keynote speakers, invited speakers and the participants to the 19th International Conference on Emerging Nuclear Energy Systems 2019 (ICENES2019). ICENES2019 is jointly organized by Institut Teknologi Bandung (ITB), Indonesia and Bahçeşehir University, Turkey, in collaboration with Karabuk University, Turkey, International Atomic Energy Agency (IAEA), American Nuclear Society (ANS), Indonesian Nuclear Society (HIMNI), Indonesian Atomic Energy Agency (BATAN), and Indonesian Atomic Energy Regulatory Body (BAPETEN). The main objective of ICENES is to provide an international scientific and technical forum for scientists, engineers, industry leaders, policy makers, decision makers and young scientists / professionals who will shape future energy supply and technology, for a broad review and discussion of various advanced, innovative and non-conventional nuclear energy production systems. The new dimension of ICENES since 2007 is to extend the forum, which will also comprise innovative non-nuclear technologies, such as hydrogen energy, solar energy, deep space exploration, etc. with emphasis on UNTHINKABLE IDEAS on sound scientific-technical basis. Earlier conferences were held in Graz (Austria), Lausanne (Switzerland), Helsinki (Finland, 1983), Madrid (Spain, 1986), Karlsruhe (Germany, 1989), Monterey (USA, 1991), Chiba (Japan, 1993), Obninsk (Russia, 1995), Tel-Aviv (Israel, 1998), Petten (The Netherlands, 2001), Albuquerque (USA, 2003), Brussels (Belgium, 2005) and İstanbul (Türkiye, 2007), Ericeira (Portugal, 2009), San Francisco (U. S. A., 2011), Madrid (Spain, 2013), İstanbul (Türkiye, 2015), Hefei (China, 2017). It has been the tradition of the ICENES conference series to select conference venues with unique features. For 2019, one of the most beautiful and attractive touristic, meetings, conferences regions of the World, Bali Island will host ICENES 2019. Being in Bali Indonesia, Ultimate in Diversity; Over 17,000 islands spreading between the pacific and Indian Ocean; More than 200 ethnic groups with over 300 spoken languages bridging the continents of Asia and Australia; welcomes international

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participants of ICENES 2019 to build relationships and expand horizons a multitude of amazing landscapes and biodiversity stretching along the equator line; this is Indonesia, a land of endless spectacular wonders! The program of ICENES2019 features 6 keynote speeches, 6 invited speeches and about 95 contributed oral presentations, which come from 22 different countries, namely: Indonesia, Turkey, Japan, USA, India, China, Iran, KSA, Lithuania, Mongolia, Brazil, South Korea, Spain, Italy, South Africa, Belgium, Russia, Bangladesh, Malaysia, Vietnam, Germany, and France. All papers will be reviewed after they are presented at this event. Selected papers will be invited to be published in the International Journal of Energy Research, the International Journal of Hydrogen Energy. and Institute of Physics Conf. Series To all participants, we hope that you will enjoy much a very short stay in Bali, the last paradise in the Earth. Finally, we wish to express our honest appreciation to all of the invited speakers and the presenters for their important contributions and also to the members of the program committee for their outstanding works in selecting abstracts and organizing the program.

Best Regards,

ICENES2019 Chairman, Prof. Abdul Waris Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia

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COMMITTEES Host organizations Institut Teknologi Bandung Bahçeşehir University President of ICENES Prof. Dr. Sümer ŞAHİN Vice-President of ICENES Prof. Dr. Hacı Mehmet ŞAHİN Conference Chairmen Prof. Dr. Abdul WARIS Prof. Dr. Zaki SUUD Technical Program Chairman Dr. Sidik PERMANA Scientific Secretaries Dr. Syeilendra Pramuditya Dr. Eng. Dwi Irwanto Standing Committee José M. MARTÍNEZ-VAL Rainer SALOMAA Jacques LIGOU Günter KESSLER Guillermo VELARDE Ralph MOIR Hideshi YASUDA Anatoloy ZRODNIKOV Louis TEPPER A. H. VERKOOIJEN Tom MEHLHORN Hamid Ait ABDERRAHIM Sümer ŞAHİN Pedro VAZ Wayne MEIER Emilio Minguez TORRES Hacı Mehmet ŞAHİN Yican WU Scientific Committee C. Rubbia, T. Nejat Veziroğlu, Mohamed Abdou, Ibrahim Dinçer, Laila El-Guebaly, Mohamed S. El- Genk, Yonghee Kim , Hiroshi Sekimoto, Jose Manuel Perlado, Kazuo Tanaka, Waclaw Gudowski, Şenay Yalçın, Mohamed Sawan, Andrey Gulevich, V. Jagannathan, Minghuang Wang, Chen Zhibin, Ruzhu Wang, Xiaojing Liu, Francesco Orsitto In cooperation with Karabuk University, IAEA, ANS, HIMNI, BATAN, and BAPETEN

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ICENES 2019 SCHEDULES

Sunday, Oct 6, 2019

Time Event 13.30 Registration At Holiday Inn Hotel (Lobby)

Monday, Oct 7, 2019

Time Event 08.00 Registration 08.50 Opening Ceremony (Cinnamon Ballroom) Plenary session 1 (Keynote Speaker) 09.10 Prof. Carlo Rubbia (Italy) 09.50 Tea / Coffee Break Plenary session 2 (Keynote Speakers) 10.00 Prof. Hiroshi Sekimoto (Japan) Prof. Yican Wu (China) Plenary session 3 (Invited Speakers) 11.00 Prof. Zaki Suúd (Indonesia) Prof. Anhar Riza Antariksawan (Indonesia) 12.00 Lunch Break (Lunch Talk) Parallel session 1 Room1 Room2 Room 3 13.30-13.45 ABS-1 ABS-40 COAUTHOR 13.45-14.00 ABS-2 ABS-51 MEETING 14.00-14.15 ABS-3 ABS-22 14.15-14.30 ABS-7 ABS-32 ABS-26 14.30-14.45 ABS-18 ABS-33 ABS-28 14.45-15.00 ABS-31 ABS-55 ABS-16 15.00-15.15 ABS-34 ABS-60 ABS-23 15.15-15.30 ABS-43 ABS-21 ABS-9 Parallel session 2 Room1 Room2 Room 3 16.00-16.15 ABS-77 ABS-61 ABS-58 16.15-16.30 ABS-14 ABS-46 ABS-65 16.30-16.45 ABS-17 ABS-82 ABS-89

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Parallel session 2 Room1 Room2 Room 3 16.45-17.00 ABS-20 ABS-52 ABS-4 17.00-17.15 ABS-19 ABS-83 ABS-25 17.15-17.30 ABS-24 ABS-15 ABS-57

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Tuesday, Oct 8, 2019

Time Event 08.00 Registration Plenary session 4 (Keynote Speakers) (Cinnamon Ballroom) 08.30 Prof. Mohamed Abdou (USA) Prof. Emilio Minguez Torres (Spain) Prof. Jose Rubens Maiorano (Brazil) 10.00 Tea / Coffee Break Plenary session 5 (Invited Speakers) 10.15 Dr. Lars Jorgensen (USA) Dr. Kurtubi (Indonesia) Plenary session 6 (Invited Speakers) 11.15 Prof. Jazi Eko Istiyanto (Indonesia) Dr. Susilo Widodo (Indonesia) 12.15 Lunch Break Parallel session 3 Room1 Room2 Room 3 13.30-13.45 ABS-47 ABS-66 ABS-53 13.45-14.00 ABS-48 ABS-36 ABS-56 14.00-14.15 ABS-75 ABS-70 ABS-59 14.15-14.30 ABS-54 ABS-37 ABS-72 14.30-14.45 ABS-91 ABS-79 ABS-5 14.45-15.00 ABS-92 ABS-80 ABS-88 15.00-15.15 ABS-67 ABS-84 ABS-90 15.15-15.30 ABS-93 ABS-71 ABS-76 15.30-15.45 ABS-35 ABS-86 ABS-85 15.45-16.00 ABS-30 ABS-87 ABS-62 Parallel session 2 Room1 Room2 Room 3 16.15-16.30 ABS-45 ABS-11 ABS-63 16.30-16.45 ABS-78 ABS-38 ABS-12 16.45-17.00 ABS-94 ABS-39 ABS-49 17.00-17.15 ABS-81 ABS-64 17.15-17.30 ABS-74

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Wednesday, Oct 9, 2019

Time Event 09.00 Cultural Events/ Bali Tours 12.15 12.15 Lunch (On Trip) Cultural Events/ Bali Tours 15.30 Return To Hotel

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CONFERENCE MAP

VENUE LAYOUT (HOLIDAY INN RESORT, BARUNA, BALI)

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TABLE OF CONTENT Preface ...... i ICENES 2019 Schedules ...... iv Conference Map ...... viii VeNUE LaYOUT ...... viii (Holiday Inn ReSORT, BARUNA, BALI) ...... viii Table of Content ...... ix [ABS-1] ...... 1 A SMALL MODULAR REACTOR FOR CONTRIES FOR MEDIUM ENERGY PRODUCTION ...... 1 Sumer ŞAHİN and Başar ŞARER ...... 1 [ABS-2] ...... 2 DEVELOPMENT OF LINKAGE PROGRAM CODE OPENMC AND ORIGEN2.2 FOR NEUTRONIC ANALYSIS AND BURNUP NUCLEAR REACTOR PROGRAM Analysis .... 2 Muhammad Ilham (a*), Helen Raflis (a), Zaki Suud (a) ...... 2 [ABS-3] ...... 3 Reflector Material Selection for Core Design of Modular Gas-cooled Fast Reactor (GFR) using OpenMC Code ...... 3 Helen Raflis (a*), Muhammad Ilham (a), Zaki Su’ud (a), Abdul Waris (a), and Dwi Irwanto (a) ...... 3 [ABS-7] ...... 4 Comparative Study on Neutronic Characteristics of VHTR core ...... 4 Odmaa Sambuu(a,b*), Khukhsuvd Batsaikhan (b), Jamiyansuren Terbish (b) Munkhbat Byambajav (a,b) ...... 4 [ABS-13] ...... 6 R&D Activities of Neutronics and Advanced Nuclear Systems at INEST • FDS Team ...... 6 Yican Wu*, FDS Team ...... 6 ix Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-16] ...... 7 The Influence of the Volume Ratio of Moderator and Fuel on the Core Size of Small and Long-life Reactor ...... 7 Yanting Sun *, Qi Yang, Jun Gao, FDS Team ...... 7 [ABS-18] ...... 8 A New Design Concept of Burnable Poison for PWR Core – BP Attached to GT ... 8 Aiman Dandi and Myung Hyun Kim* ...... 8 [ABS-23] ...... 9 DEVELOPMENT OF SMALL MODULAR LFR DESIGNS FOR ICEBREAKER SHIP ...... 9 Tung Dong Cao Nguyen, Jiwon Choe, Xianan Du, Sooyoung Choi and Deokjung Lee* ...... 9 [ABS-26] ...... 10 NUCLEAR FUEL COMPOSITION FOR REPETITIVE CLOSED FUEL CYCLE ...... 10 V. E. Moiseenko, S.V. Chernitskiy ...... 10 [ABS-25] ...... 11 MCS Monte Carlo multi-physics depletion analysis of an OPR-1000 reactor ...... 11 Vutheam Dos (a), Hyunsuk Lee (a), Jiwon Choe (a), Matthieu Lemaire (a), Ho Cheol Shin (b), Hwan Soo Lee (b), and Deokjung Lee (a*) ...... 11 [ABS-31] ...... 12 Recent Advances in Control of Nuclear Power Plants ...... 12 Areai Nuerlan and Rizwan-uddin ...... 12 [ABS-34] ...... 13 Fuzzy Logic-Based Feedwater Controller for the Super Fast Reactor ...... 13 Sutanto, Marili Santi, Anggun Dwi Lestari, Ayu Jati Puspitasari ...... 13 [ABS-42] ...... 14 ThorCon Design StatuS ...... 14 Jack Devanney(a*), Robert Hargraves(b), Lars Jorgensen(a), Ralph Moir(c) ... 14 [ABS-43] ...... 15

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Initial Core Design of Modified CANDLE Reactor with Pb 208-Bi eutectic as a coolant ...... 15 Nina Widiawati1,a), Zaki Su’ud1,b), Dwi Irwanto1,c) and Sidik Permana2,d) . 15 [ABS-47] ...... 16 Preliminary Neutronic Analysis of Small MSFR with Th-U FueL ...... 16 Abdul Waris1*, Cici Wulandari2 , Robi Dany Riupassa2, Dwi Irwanto1 , and Asril Pramutadi AM1 ...... 16 [ABS-48] ...... 17 Neutronic Analysis of HTTR 30 MWth with (Th, U) Fuel and CO2 Coolant ...... 17 Abdul Waris1*, Fauzan G. Anshari2, Anni Nuril Hidayati2, Zaki Su’ud1 ...... 17 [ABS-53] ...... 18 Investigation of the use of different Cladding Materials on the Neutronic and Thermal-hydraulic Parameters of Tehran Research Reactor ...... 18 Farhad Salari, Mohammad Reza Nematollahi *, Mohsen Ebrahimian ...... 18 [ABS-54] ...... 19 Morphology of Y-Ti Nano-oxides in ODS Alloys Irradiated with High Energy Heavy Ions ...... 19 J.H. OConnell (a*), V.A. Skuratov (b), A.S. Sohatsky (b), K. Kornieieva (b), A.D. Volkov (c), M. Zdorovets (d) ...... 19 [ABS-56] ...... 20 A THORIUM-FUEL PIN NEUTRONIC ANALYSIS USING DIFFERENT NUCLEAR CODES ...... 20 Felipe M. G. Pereira, Renato V. A. Marques, Márcia S. Santos, Carlos. E. Velasquez and Claubia Pereira ...... 20 [ABS-59] ...... 21 OVERVIEW ABOUT THE STUDIES ON ADVANCED NUCLEAR REACTORS PERFORMED AT THE FEDERAL UNIVERSITY OF MINAS GERAIS ...... 21 A. A. P. Macedo, M. Gilbert, A. L. Vieira, A. A. Cunha, G. H. P. Dias, M. C. Ramos, M. E. Scari, F. C. Silva, P. A. L. Reis, C. A. M. Silva, A. L. Costa, M. A. F. Veloso, C.E. Velasquez and C. Pereira ...... 21 xi Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-67] ...... 22 Neutronic Analysis of Sodium-Cooled Fast Reactor (SFR) Design with Various Fuel Types Using Shuffling StrategY ...... 22 Mohammad Ali Shafii, R. Septi, F. Handayani, A. Arkundato, Zaki Su’ud …… 22 [ABS-68] ...... 23 Simulation of different density mixing pebble flow in a two-dimension circulating packed bed systeM ...... 23 Dwi Irwanto, Sparisoma Viridi ...... 23 [ABS-72] ...... 24 The conceptual design of thorium-based molten salt energy amplifieR ...... 24 Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*) ...... 24 [ABS-74] ...... 25 Analysis of integrated target in the thorium-based molten salt energy amplifieR Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*) ...... 25 [ABS-75] ...... 26 INVESTIGATION OF THORIUM- ALTERNATIVE FUEL MIXTURES IN A GAS TURBINE MODULAR HELIUM REACTOR ...... 26 Sumer ŞAHİN (a), Hacı Mehmet ŞAHİN (b*), Özgür Erol (c) ...... 26 [ABS-91] ...... 27 Modified CANDLE Analsysis Using Microscopic Cross Section From SLAROM Code for Detail Analysis of Multi Scenario Modified CANDLE Scheme ...... 27 Zaki Su’ud, Fitria Miftasani, Feriska H. Irka, Nina Widiawati, Helen Raflis, H Sekimoto, Sumer Sahin, Mehmed Sahin, Zuhair ...... 27 [ABS-92] ...... 28 Preliminary Study of Nuclear Safety System Analysis and Simulation for Molten Salt ReactOR ...... 28 Muhammad Ilham (a), Cici Wulandari (a), Putranto Ilham Yazid (a), Sidik Permana (a) ...... 28 xii Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-93] ...... 29 Design Development of PeLUIt-40 a Small Cogeneration Nuclear Power PlanT . 29 Topan Setiadipura(a*), Dwi Irwanto(b), Arya Adhyaksa Waskita(a), Hery Adrial(a), Suwoto(a), Zuhair(a) ...... 29 [ABS-5] ...... 30 preliminary calculation on the containment external cooling effect for FLEX strategy using containment analysis codeS ...... 30 KYUNGHO NAM, Bum-Soo Youn ...... 30 [ABS-6] ...... 31 THORIUM IN BRAZIL ...... 31 Jose Rubens Maiorino ...... 31 [ABS-9] ...... 32 The Calculation of the Neutron Yield Distribution for the Gas Target of HINEG . 32 Z. Yang (a, b), Y. Zhang (a), Z. Wang (a), S. Chen (a*), W. Wang (a), T. Li (a), Y. Wu (a), FDS Team ...... 32 [ABS-11] ...... 33 A Coupling Control Strategy of Accelerator and Gas Target for High Intensity D-T Fusion Neutron GeneratoR ...... 33 C. Zhao(a,b), J. Wang(a), Y. Wang(a), Z. Wang(a), Q. Zhang(a), Y. Zhang(a*), FDS Team(a) ...... 33 [ABS-21] ...... 34 THE AP-Th 1000 AN ADVANCED CONCEPT TO USE MOX OF THORIUM IN A CLOSED FUEL CYCLE ...... 34 Giovanni Laranjo de Stefani; Jose Rubens Maiorino ...... 34 [ABS-22] ...... 35 Metallic fission product transport in TRISO coated particleS ...... 35 Johannes Neethling (a*), Jaco Olivier (a), Jacques OConnell (a) ...... 35 [ABS-32] ...... 36

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Thermal Behavior Study of Vertical Heater Vertical Cooler (VHVC) Natural Circulation Loop (NCL) with Horizontal Width VariatioN ...... 36 (1)(a)Duwi Hariyanto and (1)(2)(b)Sidik Permana ...... 36 [ABS-33] ...... 37 Dynamic visualization on Erosion Behavior of a Solid Plate with Particle MethoD ...... 37 M Ifthacharo(a*), A P A Mustari(b), S Permana(b), A Nuril(a) ...... 37 [ABS-38] ...... 38 Development of a New Hybrid Method to Assess the Frequency of occurrence of Loss of Offsite Power to the Nuclear Power Plants ...... 38 Shahabeddin Kamyab 1; Mohammadreza Nematollahi 1,2; Faramarz Yousefpour 3 ...... 38 [ABS-39] ...... 39 Numerical and 2D PIV Study of the Effects of Hydrodynamic Parameters on the Flow Accelerated Corrosion downstream of a Gate ValvE ...... 39 Abbas Sedghkerdar1, Mohammadreza Nematollahi 1, 2 * and Ali Erfaninia 2, 3 ...... 39 [ABS-40] ...... 40 GAMMA HEATING EVALUATION OF RSG G.A. SIWABESSY SILICIDE CORE ...... 40 Anis Rohanda1,3, Abdul Waris2, Rizal Kurniadi2, Syaiful Bakhri3 ...... 40 [ABS-41] ...... 41 The Molten Woods Metal Initial Velocity Variations Effect on Breaching ProcesS ...... 41 A N Hidayati (*a), A Waris (b), A P A Mustari (b), N A Aprianti (b), M Iftacharo (c) ...... 41 [ABS-44] ...... 42 Opportunities of the Nuclear Energy Integrated with Variable Renewables ...... 42 Emilio MINGUEZ ...... 42 [ABS-51] ...... 43

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Simulation Of TRIGA 2000 Bandung Neutron And Gamma Data Dose Using MCNPX, PHITS And Extrapolation With Tenth Value Layer (TVL ) ...... 43 Rakotovao Lovanantenaina Oméga(a), Sidik Permana(b), Rini Heroe Oetami(c), Rasito(d) ...... 43 [ABS-55] ...... 44 Preliminary Study of the Radionuclides Transport Behaviour During Normal Operation of Direct-cycle Gas Turbine HTr ...... 44 I Wayan Ngarayana 1,2; Kenta Murakami 1 ...... 44 [ABS-60] ...... 45 Grading for the Maintenance Activities of the Advanced Nuclear Reactor Using Modified Fuzzy FMEA & Expert Judgement MethodologY ...... 45 I Wayan Ngarayana 1,2; Kenta Murakami1 ...... 45 [ABS-66] ...... 46 NEUTRONIC Assesment of HTGR study design on the use of ZrC TRISO-COATED PARTICLE (TRIZO) ...... 46 Fitria Miftasani, Zaki Suud, Dwi Irwanto ...... 46 [ABS-69] ...... 47 E Wall Effect and Fast Neutron Irradiation Impact in Thermal Hydraulics Analysis of HTR-10 ...... 47 Bilal El Bari(a*), Dwi Irwanto(a) ...... 47 [ABS-70] ...... 48 Characteristics of Austenite SS 316L Steel under High Temperature Molten Lead Bismuth Environment: the Molecular Dynamics Simulation of Corrosion InhibitioN ...... 48 Artoto Arkundato(1), Fiber Monado(2), Mohammad Ali Shafii(3), Ratna Dewi Syarifah(4) ...... 48 [ABS-76] ...... 49 An Overview of the Applicability of SNI IEC 61331-1:2016 on Radiation Apron for Medical Radiation UsE ...... 49 Suzie Darmawati, Sunarto, Hanna Yasmine, Sigit Santosa ...... 49 xv Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-79] ...... 50 Corrosion Investigation of Silicon Carbide and Zirconium Carbide Caused by Silver at High Temperature: A Preliminary StudY ...... 50 Abu Khalid Rivai, Mardiyanto, Sumaryo, Bambang Sugeng, Nanda Shabrina 50 [ABS-80] ...... 51 Experimental Investigation on Corrosion Resistant of Zirconium Carbide Coating at Elevated TemperaturE ...... 51 Abu Khalid Rivai, Mardiyanto, Sumaryo, Bambang Sugeng, Nanda Shabrina 51 [ABS-84] ...... 52 Evaluation on Environment Nuclear Radiation at Various Cities in Java Island ... 52 Imam Ghazali Yasmint, Ahmad Lathiiful Quluub and Sidik Permana ...... 52 [ABS-85] ...... 53 Consideration of Intrusion Strategies and Insider-Outsider Collusion in Analyzing Physical Protection System (PPS) Using a Stochastic EASI ModeL ...... 53 Yanuar Ady Setiawan ...... 53 [ABS-86] ...... 54 Criticality Investigations on Experimental Power Reactor with Thorium-based Nuclear FueL ...... 54 Zuhair#, R.Andika Putra Dwijayanto#, Suwoto#, Zaki Su’ud* ...... 54 [ABS-87] ...... 55 Neutron Beam Characterization on the Beam Tubes of G.A. Siwabessy Reactor Using Monte Carlo MethoD ...... 55 Rasito, Zaki Su’ud, and Sidik Permana ...... 55 [ABS-88] ...... 56 Evaluation of Wind power energy potential in Dhahran, Saudi ArabiA ...... 56 Gaydaa Al Zohbi (a*), Fatima Ali Wuhayb (a) ...... 56 [ABS-90] ...... 57 STUDY ANALYSIS OF TRANSMUTATION SYSTEM USING ACCELERATOR DRIVEN SUBCRITICAL REACTOR ...... 57 xvi Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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Sudarmono (a*), Endiah Puji Hastuti (b*), Anis Rohanda (c*), Andi Sofrany Ekariansyah (d*), Y.Kasezas (e*), Suwoto (f*) ...... 57 [ABS-36] ...... 58 Neutronic analysis of HCLL-type blankets for ITER ...... 58 Indah Rosidah, Dwi Irwanto, Abdul Waris, Zaki Suud ...... 58 [ABS-37] ...... 59 Comparative of blanket reactor design in assuring of tritium self-sufficiency conditioN ...... 59 Indah Rosidah, Dwi Irwanto, Abdul Waris, Zaki Suud ...... 59 [ABS-64] ...... 60 WHAT INERTIAL FUSION R&D CAN DO FOR YOU IN SCIENCE AND TECHNOLOGY ...... 60 J. Manuel Perlado ...... 60 [ABS-71] ...... 61 STRUCTURAL MATERIAL SELECTION AND ENERGY MULTIPLICATION IN FUSION REACTORS WITH TRANSURANIUM NUCLEAR WASTE ...... 61 Sümer Şahin (a*), Hacı Mehmet Şahin (b), Hüseyin Şahiner (c), Güven Tunç (d) ...... 61 [ABS-89] ...... 63 Safety Assessment of Wendelstein 7-X Experimental Nuclear Fusion Facility in the Case of Coolant Ingress into Vacuum VesseL ...... 63 E. Ušpuras, A. Kaliatka, T. Kaliatka ...... 63 [ABS-4] ...... 64 Noise Experiments in BRAHMMA Subcritical System using Isotopic Poisson Source and Accelerator based Neutron SourcE ...... 64 Nirmal Kumar Ray(a,b*), Rajeev Kumar(c), TarunPatel(a), P.S. Sarkar(a,b), L. M. Pant(b,d) ...... 64 [ABS-15] ...... 66

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Neutron sources for Fusion-Fission hybrid reactors based on magnetic plasma confinement : analysis of design parameters and technology readiness levelS .. 66 Francesco Paolo Orsitto ...... 66 [ABS-20] ...... 67 COMPARISION OF NEUTRON FLUX DISTRIBUTION OF 233UO2, 235UO2, (TH- 233U)O2 AND (TH-235U)O2 FUEL IN THE ACCELERATOR DRIVEN SUBCRITICAL REACTOR ...... 67 Nguyen Mong Giao (a), Tran Minh Tien (b*) ...... 67 [ABS-25] ...... 69 BURNING OF MINOR ACTINIDES IN STELLARATOR-MIRROR FUSION-FISSION HYBRID ...... 69 S.V. Chernitskiy (a), V.E. Moiseenko (a), O. Agren (b) ...... 69 [ABS-57] ...... 70 Research and Development on ADS for Minor Actinides Transmutation and Thorium Utilization ...... 70 Marcia S. Santos(1), Renato V. A. Marques(1), Felipe Martins(1), Graiciany de P. Barros(2), Carlos E. Velasquez(1), Antonella L. Costa(1), Maria Auxiliadora F. Veloso(1) and Claubia Pereira(1) ...... 70 [ABS-58] ...... 71 Proposal of a Fusion-Fission System based on the ITER ProjecT ...... 71 Renato V. A. Marques, Marcia S. Santos, Felipe Martins, Carlos E. Velasquez and Claubia Pereira ...... 71 [ABS-61] ...... 72 Case study of a nuclear-solar PV hybrid base load planT ...... 72 A. K. Mathur, V. Jain, S. A. Khan, V. Jagannathan and Suneet Singh ...... 72 [ABS-65] ...... 73 Fusion-fission hybrid system experiment based on a TRIGA reactor in a sub- critical configuration ...... 73 Fabio Panza, Mario Carta, Marco Ciotti, Nadia Cherubini, Alessandro Dodaro, Valentina Fabrizio, Luca Falconi, Francesco Filippi, Luigi Lepore, Francesco xviii Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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Orsitto, Mikhail Osipenko, Mario Palomba, Giovanni Ricco, Marco Ripani, Massimo Salvatores ...... 73 [ABS-77] ...... 74 Coupling Electron Accelerators to Heavy Water Reactors: Photon Transport and Photoneutron Production ...... 74 Silvia da Costa Frias de Barros1, Gitae Kim2, Yeseul Seo2, Douglas A. Fynan274 [ABS-14] ...... 75 Climate and Health Benefits from the Expansion of Nuclear PoweR ...... 75 Minghuang Wang, Xuewei Fu, Chao Lian, Dehong Chen, Zhibin Chen, Yunqing Bai, Fang Wang, Liqin Hu, Yican Wu*, FDS Team ...... 75 [ABS-17] ...... 76 Development of Hexagonal-Z Geometry Capability in RASTK for Fast Reactor Analysis ...... 76 Tuan Quoc Tran, Alexey Cherezov, Xianan Du, Jinsu Park, Deokjung Lee* ...... 76 [ABS-19] ...... 77 Phased and Dynamic Failure Mechanisms Considered Reliability Analysis of the Diesel Generator System after LOOP ...... 77 Daochuan Ge, Zhen Wang, Shanqi Chen, Chao Chen, Zhibin Chen*, Fang Wang, FDS Team ...... 77 [ABS-24] ...... 78 Tools to Educate, and Dispel Myths and MisconceptionS ...... 78 Zahra Hanifah and Rizwan-uddin ...... 78 [ABS-30] ...... 79 The Development of Miniature Nuclear Reactor Props as a Learning MediA ..... 79 Casmika Saputra (a), Abdul Waris (b*) ...... 79 [ABS-45] ...... 80 Multilateral Approach to the Nuclear Fuel Cycle: An Alternative and the Malaysian Perspectives ...... 80 Bashillah Baharuddin, Abdul Muin Abdul Rahman and Siti A’iasah Hashim ... 80

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[ABS-78] ...... 81 Energy for our future: a synergetic approach between short-term nuclear fission and long-term nuclear fusioN ...... 81 J.-M. Noterdaeme ...... 81 [ABS-94] ...... 83 WHY INDONESIA SHOULD DEVELOP NUCLEAR POWER? ...... 83 Muhammad Busyairi ...... 83 [ABS-12] ...... 84 Intelligent Fault Diagnosis Method and Computer Aided Regulation System in Beam Commission of HINEG FacilitY ...... 84 C. Zhao(a,b), J. Wang(a), Y. Wang(a), Z. Wang(a), Q. Zhang(a), Y. Zhang(a*), FDS Team(a) ...... 84 [ABS-35] ...... 85 STUDY HEAT TRANSFER IN NATURAL CIRCULATION OF LIQUID SODIUM FOR STEADY STATE AND TRANSIENT CONDITIONS ...... 85 Rindi Wulandari(1); Sidik Permana(2); Suprijadi (3) ...... 85 [ABS-46] ...... 86 Neutronic Design of 100 MWe MSR with Th-Pu-MA FueL ...... 86 Cici Wulandari(a) , Abdul Waris(b*), Robi Dany Riupassa(a), Sidik Permana(b) , and Yazid Bindar(c) ...... 86 [ABS-49] ...... 87 Comparative Study of Different Monte Carlo Methods for Multi-Group Neutron Diffusion Constants Generation ...... 87 Fatemeh Mohammadhasani, Mohammadreza Nematollahi* ...... 87 [ABS-52] ...... 88 ANALYSIS OF NEUTRONIC AND NON-PROLIFERATION ASPECTS WHIT FUEL VARIATIONS ON MOLTEN SALT REACTOR (MSR) BASED ON THORIUM FUEL ..... 88 Ade Maulana Fadillah(a*), Sidik Permana (b) ...... 88 [ABS-62] ...... 89

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Experimental and Numerical Study of Flow Accelerated Corrosion in a T-junction By Utilizing CFX and 2D PIV ...... 89 Ali Erfaninia(a,b,*) , Mohammadreza Nematollahi(a,b) ...... 89 [ABS-63] ...... 90 2D PIV Study of Flow Accelerated Corrosion Downstream a Typical Industrial Gate Valve ...... 90 Abbas Sedghkerdar(a), Ali Erfaninia(a,b,*), Mohammadreza Nematollahi(a, b) ...... 90 [ABS-81] ...... 91 Ag and Pd Fission Product Implantation on SiC layer in TRISO Fuel Particle of HTGR using SRIM/TRIM Monte Carlo Computer CodE ...... 91 Mardiyanto Mangun Panitra1 and Abu Khalid Rivai1 ...... 91 [ABS-82] ...... 92 Investigation on Physical Interaction of Ag and Pd Fission Products with ZrC layer in TRISO Fuel Particle of HTGR using SRIM/TRIM Monte Carlo Computer Code . 92 Mardiyanto Mangun Panitra1 and Abu Khalid Rivai1 ...... 92 [ABS-83] ...... 93 Modeling of Drainage System on Modified Freeze Valve in a Molten Salt ReactoR ...... 93 Robi Dany Riupassa (a*), Abdul Waris (b), Khairul Basar (b), Novitrian (b), Yazid Bindar (c), Cici Wulandari (a) ...... 93

xxi Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-1] A SMALL MODULAR REACTOR FOR CONTRIES FOR MEDIUM ENERGY PRODUCTION Sumer ŞAHİN and Başar ŞARER

1) Bahcesehir University, Faculty of Engineering and Natural Sciences, BeÅŸiktaÅŸ, İstanbul, TURKİYE Email: [email protected] 2) Near East University, LefkoÅŸa/KKTC, Turkish Republic of Northern Cyprus, Mersin 10, TURKİYE ABSTRACT

The Fixed Bed Nuclear Reactor (FBNR) is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on Small Reactors (70-120 MWel) without On-site Refueling. The FBNR is adequate for developing countries with small electric grids and limited investment capabilities, as well as with the weakness of manpower for development of nuclear power plants. The FBNR fuel element consists of 500 microns in diameter UO2 micro spheres covered by 25 microns thick zirconium cladding embedded in a spherical zirconium matrix with 300 μm thick Zircaloy-4 cladding to form a 15 mm diameter fuel element. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and diameter of 170 cm by upwards flowing light water coolant with 280 oC entry and 340 oC exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator. FBNR can operate both with conventional nuclear fuel 235U as well as with 233U and make use of the worldwide abundant thorium on that way. Utilization potential of alternative fuels, such as thorium, reactor grade plutonium and minor actinides enables FBNR to have a high level of sustainability. In the present work, 9 % enriched UO2 fuel is used. Calculations are conducted with the MCNPX 2.7 code. As, it is not practical to tackle millions of fuel elements in the same run, at first unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5 % range, i.e., on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 20 is selected for further investigations. Temporal variation of the reactor criticality is pursued for a reactor power of 400 MWth (~ 120 MWel net) with keff = 1.311 at the beginning of life down to keff = 1.044 for a full power operation time of 570 days, leading to a fuel burn up of 44 GWd/MTU.

Keywords: nuclear energy; small modular reactor; fixed bed nuclear reactor; uranium Topic: Advanced Fission Systems

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ICENES 2019 Program Book

[ABS-2] DEVELOPMENT OF LINKAGE PROGRAM CODE OPENMC AND ORIGEN2.2 FOR NEUTRONIC ANALYSIS AND BURNUP NUCLEAR REACTOR PROGRAM ANALYSIS

Muhammad Ilham (a*), Helen Raflis (a), Zaki Suud (a)

Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, INDONESIA E-mail: [email protected]

ABSTRACT

This thesis research discusses the development of program code that coupling the ORIGEN2.2 for burn-up analysis with the Monte Carlo program, OpenMC, for neutron analysis program called OpenMC-ORIGEN (Op-OR). The results of this program have been compared with benchmark results from previously published results and MCNP6 program. The acquired results show a good agreement with benchmark results. This program is written using the Python3 program language. This linkage program codes perform well for designing a new advanced reactor and analyzing the neutronic parameter and burnup/depletion calculation.

Keywords: OpenMC, Monte Carlo, ORIGEN, Burnup, Parallelization, Op-OR

Topic: Advanced Fission Systems

2 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-3] REFLECTOR MATERIAL SELECTION FOR CORE DESIGN OF MODULAR GAS- COOLED FAST REACTOR (GFR) USING OPENMC CODE

Helen Raflis (a*), Muhammad Ilham (a), Zaki Su’ud (a), Abdul Waris (a), and Dwi Irwanto (a)

1) Department of Physics, Faculty Science and Technology of Universitas Islam Negeri Sunan Gunung Djati Bandung, Indonesia *[email protected] 2) Bolabot Techno Robotic Institute of CV. Sanjaya Star Group, Bandung, Indonesia 3) Department of Mechanical Engineering, Universitas Muhammadiyah Tasikmalaya, Indonesia

ABSTRACT

The selection of reflector material for core design of modular Gas-cooled Fast Reactor (GFR) that has the potential for use actinide recycling and closed fuel cycle using helium gas as the main coolant, high working temperature, and low void reactivity effect has been concerned. The modular GFR has the ability to maintain the criticality to achieve a long life of the fast reactor. The neutron economy should be very good to improve the neutron economy via reduced neutron leakage so that the long life condition achieved. In this research, the selection of potential reflector in modular GFR investigated from physics phenomenon and criticality condition. The best neutron reflector is essential to maintain the neutron economy which fast reactor design usually has a higher neutron leakage. The various alternative reflector materials such as pure lead, pure nickel, pure magnesium, pure bismuth, Ba2Pb, BeO, SiC, PbO, Zr3Si2, ZrS2, and SS-HT9 are calculated investigated from the neutronics perspectives. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VII.b5) nuclear data and continuous energy of OpenMC code. The physics parameters characterized including the value of keff, reflecting performance, core neutron spectrum, core leakage, power distribution, and neutron flux profiles to understand the behavior of each reflector material. The most important neutronic parameters of GFR core design is determined for the beginning of life (BOL) conditions. Finally, we got the result that lead-based reflectors and zirconium- based-reflectors are good reflector candidates for the design of modular GFR.

Keywords: Modular Gas-cooled Fast Reactor, Reflector, Monte Carlo Method, Core Leakage.

Topic: Advanced Fission Systems

3 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-7] COMPARATIVE STUDY ON NEUTRONIC CHARACTERISTICS OF VHTR CORE

Odmaa Sambuu(a,b*), Khukhsuvd Batsaikhan (b), Jamiyansuren Terbish (b) Munkhbat Byambajav (a,b) a) Department of Chemical and Biological Engineering, School of Engineering and Applied Sciences, National University of Mongolia, Ikh surguuliin gudamj 3, Sukhbaatar district, Ulaanbaatar 14201, Mongolia (*[email protected]) b) Nuclear Research Center, National University of Mongolia, Peace Avenue 122, Bayanzurkh District, Ulaanbaatar 14201, Mongolia

Abstract

The solid and annular cylindrical prismatic core designs of VHTR which were fuelled with an advanced TRISO particle fuel with additional ZrC were proposed and the preliminary neutronic analyses were carried out previously [1]. The preliminary neutronic results were compared between the cores operating at temperature of 850oC with and without ZrC additional layer and it showed that the effective neutron multiplication factor in BOC and discharged burnup was increased, while core lifetime was reduced due to existence of the ZrC layer. Neutronic feature of an annular prismatic VHTR core with ZrC-containing TRISO fuel was improved for long term operation as higher fuel burnup in effect of inner reflector.In this study, a comparative study has been performed to evaluate the capability of the alternative fuel kernel for annular prismatic VHTR core with TRIZO fuel. A representative VHTR core with different types of fuel was the same as 100 MWt at 850oC operating temperature. (U,Th)O2 and UCO kernels, as representatives of thorium, carbide fuels, are loaded into annular VHTR core. Uranium enrichment contents are adjusted to 20% for these fuel cases respectively, aiming at the comparison of oxide fuel of our previous research [1]. The neutronic analyses are performed using continuous energy Monte Carlo code MVP2.0 [2] and MVPBURN [3] with JENDL4.0 nuclear data library [4]. Based on neutronic analyses results of the present study and comparing to other fuel cases, carbide fuel case was proved to be able to provide the longest margin to the working limits, concerning the effective neutron multiplication factor at BOC, core operating lifetime and fuel burnup. This enables the carbide fuel to be competitive candidate for the VHTR usage. References: 1. Sambuu Odmaa. Batsaikhan Khukhsuvd, Terbish Jamiyansuren, Byambajav Munkhbat and Nanzad Norov, Preliminary neutronic analyses on VHTR core design, International Journal of Nuclear Safety and Simulation, Vol.9, No.2 December 2018. 2. Yasunobu Nagaya, Keisuke Okumura, Takamasa Mori, Masayuki Nakagawa, MVP/GMVP II: General Purpose Monte-Carlo Code for Neutron and Photon Transport Calculations Based on Continuous Energy and Multigroup Methods, JAERI-1348, Japan: Japan Atomic Energy Research Institute, 2005.

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3. Keisuke Okumura, Yasunobu Nagaya, Takamasa Mori, MVP-BURN Users Manual, Japan: Atomic Energy Agency, 2005. 4. Keiichi Shibata, Osamu Iwamoto, Tsuneo Nakagawa, Nobuyuki Iwamoto et al, Japanese Evaluated Nuclear Data Library-JENDL-4.0, A New Library for Nuclear Science and Engineering, J.Nucl.Sci.Tech, 2011, 48(1):1-30.

Keywords: VHTR, advanced TRISO fuel, thorium oxide fuel, carbide fuel, neutronic analysis Topic: Advanced Fission Systems

5 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-13] R&D ACTIVITIES OF NEUTRONICS AND ADVANCED NUCLEAR SYSTEMS AT INEST • FDS TEAM

Yican Wu*, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China

Abstract

Advanced nuclear energy systems have attracted more and more attention all over the world for its great superiority of sustainability, safety, and economics. The Institute of Nuclear Safety Technology, Chinese Academic of Sciences • FDS Team (INEST • FDS Team) has focused on the fundamental and applied research in the area of neutronics and advanced lead-based reactors. Neutronics is the key basis for the innovative development of nuclear systems and nuclear safety. INEST • FDS Team have been focused on the research of neutron transport physics and technology, including the development of advanced theories and software for neutron transport, technologies for neutron control and measurement, and the nuclear design and safety evaluation of advanced nuclear systems, etc. In this contribution, the latest development of Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation (SuperMC) and High Intensity D-T Fusion Neutron Generator (HINEG) are introduced. Lead-based reactor is one of the most promising nuclear systems for Generation-IV reactor, SMR(Small modular reactor) and Accelerator Driven subcritical System (ADS). INEST • FDS Team have placed more emphases on the design and R&D of China LEAd- based Reactor (CLEAR) for more than 30 years. In this contribution, the latest progress on the designs and R&D activities for CLEAR series reactors are introduced.

Keywords: Advanced nuclear systems, Neutronics, Lead-based reactor, CLEAR

Topic: Advanced Fission Systems

6 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-16] THE INFLUENCE OF THE VOLUME RATIO OF MODERATOR AND FUEL ON THE CORE SIZE OF SMALL AND LONG-LIFE REACTOR

Yanting Sun *, Qi Yang, Jun Gao, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China

Abstract

As one of the key physical parameters for the core design of small and long-life reactor, the neutron spectrum affects the core size and the reactivity temperature coefficients directly. From the view of core size optimization design, designs based on the thermal neutron spectrum and fast neutron spectrum have their advantages and disadvantages. Based on the proposed core model with annular channel fuel elements, the influence of the volume ratio of moderator and fuel on the core size was studied by using Super Multi-function Calculation Program for Nuclear Design and Safety Evaluation (SuperMC). Also the optimum volume ratio of moderator and fuel which corresponds to the minimum core size under different energy output demands (power × refueling period) was explored. Moreover, the influence of the volume ratio of moderator and fuel on the reactivity feedback was analyzed. The results show that when the reactor total energy output demand is greater than 275 MWt·Year, the core designed with a fast neutron spectrum (without moderator) has minimum size. When the reactor total energy output demand is less than 275 MWt·Year, the core designed with a thermal neutron spectrum (with moderator) has the minimum size. The less the energy output demand is, the higher volume ratio of moderator and fuel for minimum core size should be. At the same time, the volume ratio of moderator and fuel should be less than 3.18 to ensure the negative reactivity temperature coefficients. The research results in this paper will provide references for the physical design of small and long-life reactor.

Keywords: Small and long-life reactor, Neutronics, Neutron spectrum, Volume ratio of moderator and fuel

Topic: Advanced Fission Systems

7 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-18] A NEW DESIGN CONCEPT OF BURNABLE POISON FOR PWR CORE – BP ATTACHED TO GT

Aiman Dandi and Myung Hyun Kim*

Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do, Korea, 17104 *[email protected]

Abstract

A longer cycle option in Pressurized Water Reactor (PWR) core offers many benefits both in economics and waste concerns. However, there is a requirement to have stronger burnable poison rods in order to compensate increased excess reactivity. In this study, a new BP design concept called Burnable poison Attached to Guide tube (BAG) is presented. Despite similar concept is proposed in the previous studies, BAG design is unique in application. 16×16 Combustion Engineering (CE) design model is chosen as a reference for fuel assembly design. For the feasibility study of the longer cycle PWR, fuel enrichment were increased to 6.96w/o for base fuel pins and 4.10w/o for zoning pins. In this study, assembly calculation was done by DeCART-2D code without checking impact to the core design. Although BAG design cannot reduce the initial excess reactivity into reasonable level by itself, its good properties make it a very good option to give support to any conventional BP by combining it with them. Even though BAG+Erbia case have about half of the Erbia pins than that in the Erbia only case, both cases control the excess reactivity with almost the same performance. These two cases have the ability to control the excess reactivity longer than any other cases. The residual reactivity penalty of BAG+Erbia case is very low due to the fully depletion of B-10 in BAG design and a very low number of Erbia pins. The last two points are very important to design PWR core with longer operation cycle. Lastly, BAG+Erbia case provides reasonable values of power peaking factor and moderator temperature coefficient (MTC).

Keywords: Burnable Poison, PWR, Design Concept, attached to Guide Tube

Topic: Advanced Fission Systems

8 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-23] DEVELOPMENT OF SMALL MODULAR LFR DESIGNS FOR ICEBREAKER SHIP

Tung Dong Cao Nguyen, Jiwon Choe, Xianan Du, Sooyoung Choi and Deokjung Lee*

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST- gil, Ulju-gun, Ulsan, 44919, Republic of Korea [email protected], [email protected], [email protected], [email protected], [email protected]* Abstract

A conceptual development for the long-life small modular lead-bismuth eutectic fast reactor (SMLFR) has been presented in this work. The key design constraint for this fast reactor is the transportation capability in spent nuclear fuel (SNF) cask so as to be able to use as a single or cluster power propulsion for icebreakers. Another innovated feature of this suggested SMLFR is all the core components are included within a small reactor vessel, which can be immediately transferred into the SNF cask after its entire operation time. The thermal power of the SMLFR is 37.5 MW with an assumption of 40% thermal efficiency by using an advanced energy conversion system based on supercritical carbon dioxide (S-CO2) as the working fluid. It is also designed to target more than 30 years of cycle length without refueling and a small reactivity swing by adopting a breed and burn concept. For such a long-life, small and portable reactor, an excellent neutron economy is a vital requirement. A recent study has been reported that the LBE cooled fast reactor demonstrates a better performance in neutron economy, burn-up reactivity swing, and void coefficient rather than sodium fast reactor (SFR). In addition, uranium nitride (UN) with a high thermal conductivity and a high-concentrated amount of fissile fuel is chosen as one of the primary fuel candidates for LFR due to better compatibility with the LBE coolant and providing an immense improvement in neutron economy compared to uranium oxide fuel. The core inlet and outlet temperatures are 300oC and 400oC, respectively. The 15-15Ti stabilized steel is selected as cladding and structure material due to its excellent swelling resistance and stability in LBE. The performance in design and analyses of this core are conducted with the fast reactor analysis code system MC2-3/TWODANT/REBUS-3 developed by Argonne National Laboratory (ANL) and the UNIST in-house Monte Carlo code MCS with ENDF/B-VII.0 cross-section library. It is confirmed through depletion calculations that the designed reactor is capable to operate for more than 40 years without refueling and a reactivity swing less than 500 pcm. In addition, core performance features are analyzed for criticality, radial and axial power profiles and thermal-hydraulic (T/H) calculation. A preliminary T/H calculation is achieved by a T/H one-dimensional module using single-phase closed-channel model. Pin-by-pin temperature profiles are obtained as receiving the pin-wise power profiles from MCS. It is basically confirmed the outlet coolant and maximum fuel temperatures and the coolant flow velocity are within the acceptance criteria. The SMLFR core is also evaluated in view of various significant safety parameters, including control rod worth, fuel temperature coefficient, and coolant density coefficient.

Keywords: small modular reactor, fast reactor, long-cycle, icebreaker, breed and burn. Topic: Advanced Fission Systems 9 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019 Holid ICENES 2019 Program Book

[ABS-26] NUCLEAR FUEL COMPOSITION FOR REPETITIVE CLOSED FUEL CYCLE

V. E. Moiseenko, S.V. Chernitskiy

National Science Center Kharkiv Institute of Physics and Technology, Kharkiv, Ukraine

Abstract

A uranium based nuclear fuel and closed fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a balanced fuel only 238U content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. In the fuel balance calculations 9 isotopes of uranium, neptunium, plutonium and americium are used. The model accounts for fission, neutron capture and decays. Using MCNPX numerical code the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components.

Keywords: fast reactor, MCNPX calculations, neutron spectrum, reaction rates

Topic: Advanced Fission Systems

10 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-25] MCS MONTE CARLO MULTI-PHYSICS DEPLETION ANALYSIS OF AN OPR- 1000 REACTOR

Vutheam Dos (a), Hyunsuk Lee (a), Jiwon Choe (a), Matthieu Lemaire (a), Ho Cheol Shin (b), Hwan Soo Lee (b), and Deokjung Lee (a*)

a) Department of Nuclear Engineering, Ulsan National Institute of Science and Technology 50 UNIST-gil, Ulsan 44919, Republic of Korea Vutheam Dos (a): [email protected] Deokjung Lee (a*): [email protected] b) Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI) Daejeon 34101, Republic of Korea

Abstract

The Monte Carlo code MCS is developed at Ulsan National Institute of Science and Technology for the purpose of multi-physics high-fidelity analysis – neutron transport coupled with thermal/hydraulic and fuel performance (N/TH/FP) solvers - of large-scale pressurized water reactors (PWRs). In this work, the 3-D whole-core pin-wise depletion analysis of the OPR-1000 reactor system is conducted with two multi-physics coupled systems: Monte Carlo MCS with 1-D closed-channel T/H feedback (MCS/TH1D) and Monte Carlo MCS coupled with the fuel performance code FRAPCON (MCS/FRAPCON). The multi-physics coupling capability and accuracy of MCS is demonstrated by comparison of the calculated results against measured data for several neutronic parameters of interest: the critical boron concentration (boron letdown curve) and axial/radial power distributions. The solutions obtained by the MCS-based coupled systems show good agreement with the measurement data. This study therefore brings validation elements of the capability of MCS to perform the high-fidelity multi-physics simulation of a practical PWR core.

Keywords: Multi-Physics; MCS; whole-core pin-wise depletion; OPR-1000 PWR Core

Topic: Advanced Fission Systems

11 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-31] RECENT ADVANCES IN CONTROL OF NUCLEAR POWER PLANTS

Areai Nuerlan and Rizwan-uddin

School of Nuclear Science and Technology, Xi’an Jiaotong University, No.28, Xianning West Road, Xian, 710049, P.R. China

Department of Nuclear, Plasma, and Radiological Engineering Urbana, IL 61801, USA ,University of Illinois at Urbana-Champaign, 104 S Wright St

Abstract

Several new nuclear reactor designs with enhanced safety features are in either design stage or the first of the kind are being constructed. These include AP1000, CPR1000 and multi-module plants with more than one small or medium sized power units. The safety features of these have received increased attention since the Fukushima Daiichi accident. In addition to new types of cores, new control systems for core and U-tube steam generators (UTSG) are also being developed for some of these reactor designs. These designs are also expected to have better load following capabilities than GEN-II reactors. Control systems regulate the core power as well as axial power distribution (namely axial offset), possibly under load following conditions. Steam generator water level is another important variable that needs an effective control mechanism. The water level on the secondary side of UTSG must be maintained at the desirable value to ensure proper heat transfer from the reactor coolant to the secondary side, thus ensuring satisfactory operation of steam-drying equipment. Due to the shrink and swell phenomena taking place in the UTSG, the dynamic model of the UTSG is very critical for studying steam generator behavior, and thus for the design of feedwater controller. In this paper, we will review some of the recent advances made in the modeling and control of reactor cores and steam generators. Modeling approaches—such as the point reactor model, nodal reactor core model, and lumped parameter dynamic model for reactor core, and two phase flow model for the UTSG—will be reviewed. Power control methods for reactor cores, such as the feedback control with a state observer, will be reviewed. Load following control techniques for reactor cores—such as Mode A, Mode G, Mode T, Mechanical Shim and advanced multivariable frequency control methods—will also be discussed. As for UTSG level control, methods such as intelligent virtual reference feedback tuning and output feedback dissipation will be reviewed. Suggestions will be made for novel control techniques for new reactor designs.

Keywords: NPP, control, load following

Topic: Advanced Fission Systems

12 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-34] FUZZY LOGIC-BASED FEEDWATER CONTROLLER FOR THE SUPER FAST REACTOR

Sutanto, Marili Santi, Anggun Dwi Lestari, Ayu Jati Puspitasari

Department of Nuclear Technophysics, Polytechnic Institute of Nuclear Technology, National Nuclear Energy Agency, Indonesia Babarsari Street, PO BOX 6101 YKBB, Yogyakarta, Indonesia [email protected]

Abstract

A Generation IV reactor of supercritical-pressure light water fast reactor is expected to have significant improvements on the efficiency, safety and economy. Absence of boiling phenomenon with high heat capacity of the coolant at supercritical pressure results a compact plant system with low coolant flow rate and high outlet temperature, giving an increase of its efficiency up to 44%. The high outlet temperature with low coolant flow, however, leads the plant to be more sensitive to a perturbation of power to flow ratio. A change of outlet coolant temperature must be minimized to maintain the structure integrity of the reactor, such as the outlet nozzles and the main steam line. A fuzzy logic- based feedwater controller was applied to suppress the outlet coolant temperature deviation during an experience of the perturbation. Two signals of the temperature and the power changes were taken as the inputs of the fuzzy controller, and change of the feedwater coolant flow was taken as the control output. Fuzzifications used triangular member functions with Tsukamoto fuzzy model for inferencing process. Two control parameters of member function spread and input values which lead to fuzzy degree of 1 were optimized to satisfy the criteria of allowable outlet coolant temperature and stability. Performance of the control system showed that the outlet coolant temperature deviation could be kept the outlet coolant temperature within the criteria when perturbations of power to flow ratio were applied.

Keywords: Generation IV reactor, supercritical pressure, fast reactor, fuzzy control, outlet temperature

Topic: Advanced Fission Systems

13 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-42] THORCON DESIGN STATUS

Jack Devanney(a*), Robert Hargraves(b), Lars Jorgensen(a), Ralph Moir(c)

(a) ThorCon, 242 Lyons Rd, Stevenson WA 98648 USA *thorconpower.com (b) Osher Institute at Dartmouth College, Hanover NH 03755 USA (c) Vallecitos Molten Salt Research, 607 E. Vallecitos Rd, Livermore CA 94550 CA USA

Abstract

Thorium. ThorCon is a thorium converter, a denatured molten salt reactor that converts some thorium to uranium-233 and fissions it in situ. Fissile fuel is 19.75% LEU supplemented by the thorium that generates a quarter of the power, proportionately reducing costs for U-235. Thorium strengthens nonproliferation, substantially diluting chemically similar plutonium. Thorium additions are used to control reactivity via neutron absorption.Design status. ThorConIsle is a complete basic design of a fission power plant generating 500 MWe from two 557 MWt reactors. It is integrated within a large ship hull, to be towed to a shallow-water site, ballasted down, then connected to the power grid. Such a hull resting on the seabed creates new opportunities to deal with seismic motions.Power flows through fuel salt, clean salt, solar salt, then steam loops to a supercritical steam-turbine/generator. Fuel salt 704°C temperature enables 45% power conversion efficiency. ThorCon relies on existing materials and technologies to avoid R&D delays and minimize costs. It is designed in 150 to 500 ton blocks for economic shipyard manufacturing and assembly. World shipyards have sufficient capacity to manufacture 100 1-GWe power plants annually.Safety. ThorCon’s negative thermal coefficient of reactivity exceeds 4.5 pcm/°K throughout the fuel salt life. The molten salt thermal mass and a 700°C margin to boiling enable effective radiative cooling from the fuel salt drain tank. Independent computational models show loss of primary heat path and simultaneous failure of all three shutdown rods results only in acceptable creep of primary loop materials. ThorCon’s design incorporates three, separate-technology, decay-heat removal paths. ThorCon implements passive safety though physical principles rather than auxiliary safety-grade electronic or electrical control systems. At least three barriers protect against radioactive material releases from equipment failures, power blackouts, control system errors, or deliberate operator malfeasance. Our paper will describe the current status of the design and discuss ongoing multi-physics and transient modeling efforts, including seismic response of the hull, fission reactor, molten salt loops, and steam turbine-generator platform.

Keywords: molten salt, thorium, passive safety, transient modeling, seismic response Topic: Advanced Fission Systems

14 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-43] INITIAL CORE DESIGN OF MODIFIED CANDLE REACTOR WITH PB 208-BI EUTECTIC AS A COOLANT

Nina Widiawati1,a), Zaki Su’ud1,b), Dwi Irwanto1,c) and Sidik Permana2,d)

1Nuclear and Biophysics Department, Institut Teknologi Bandung, Indonesia. 2Nuclear science and engineering, Institut Teknologi Bandung, Indonesia.

Abstract

Modified CANDLE conceptual design reactors can directly consume natural uranium as a fuel input without enrichment. At first, the core was divided into several regions with the same volume. Region 1 contains natural uranium, after ten years of burnup then moved on to region 2, fuel in region 2 moved on to region 3, and so on. Whereas fuel in region 10 was removed from the core. This scheme can also make a reactor long-lived with a 10- year refuelling period. The use of Pb208-Bi as coolant also contributed to the increase in k-eff value because the cross-section absorption neutron of Pb208 is the lowest compared to the other isotopes and natural Pb itself. Some of the benefits obtained from the conceptual design reactor led to the need to prepare a Modified CANDLE initial core with easily available material. In this study, a neutronic study will be carried out on the initial core of Modified CANDLE conceptual design. Calculations were carried out using SRAC 2006. The PIJ module was used to calculate the fuel pin burnup while the CITATION module was used to calculate the multigroup diffusion of the reactor core. The nuclear library data used is JENDL 4.0.

Keywords: Modified CANDLE, Initial core, Pb208-Bi Eutectic, Natural uranium

Topic: Advanced Fission Systems

15 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-47] PRELIMINARY NEUTRONIC ANALYSIS OF SMALL MSFR WITH TH-U FUEL

Abdul Waris1*, Cici Wulandari2 , Robi Dany Riupassa2, Dwi Irwanto1 , and Asril Pramutadi AM1

1Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA 2Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA

Abstract

Molten Salt Reactor (MSR) is one of the six Generation IV of nuclear reactor systems. This reactor system MSR has many merits, such as better safety features, high thermal efficiency, and capability for waste. It was believed that the Generation IV Small Modular Reactors (SMR) are very promising for electricity generation in the middle part and the east part of Indonesia. In this study, the neutronic analysis of the Small Molten Salt Fast Reactor (MSFR) with the fuel composition is LiF-BeF2-ThF4-UF4 has been conducted. Neutronic calculations were performed by employing PIJ and CITATION modules of SRAC 2006 code with JENDL 4.0 nuclear data library.

Keywords: Small Modular Reactor, MSFR, CITATION, SRAC, JENDL 4.0

Topic: Advanced Fission Systems

16 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-48] NEUTRONIC ANALYSIS OF HTTR 30 MWTH WITH (TH, U) FUEL AND CO2 COOLANT

Abdul Waris1*, Fauzan G. Anshari2, Anni Nuril Hidayati2, Zaki Su’ud1

1Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA 2Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA

Abstract

Original HTTR (high temperature test reactor) is a 30 MWth HTGR (High Temperature Gas Reactor) with a graphite moderator, helium gas coolant, UO2 fuel with outlet coolant temperature of about 900oC. In this study, we have performed the neutronic analysis of 10 MWth HTTR with (Th, U) fuel and CO2 gas coolant. The burnup period is 1100 days, which corresponds to 3 years of fuel cycle length. The reactor calculation was performed by employing PIJ and CITATION modules of SRAC 2006 code, with the nuclear data library was derived from JENDL4.0. Several results on neutronics aspects will be presented in full paper and conference presentation.

Keywords: HTTR, Thorium, Uranium, CO2, SRAC, JENDL 4.0

Topic: Advanced Fission Systems

17 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-53] INVESTIGATION OF THE USE OF DIFFERENT CLADDING MATERIALS ON THE NEUTRONIC AND THERMAL-HYDRAULIC PARAMETERS OF TEHRAN RESEARCH REACTOR

Farhad Salari, Mohammad Reza Nematollahi *, Mohsen Ebrahimian

Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, 71936-16536 Shiraz, Iran

Abstract

In this study, the effect of cladding material on the neutronic and thermal-hydraulic parameters of the Tehran research reactor (TRR) is investigated. First, calculations are done for the first core configuration of TRR with aluminum fuel cladding. Then, using stainless steel 316, zircaloy 4, SiC, and the alloy composition of FeCrAl for the cladding calculations are performed. The considered neutronic parameters are multiplication factor, average power distribution, thermal and fast neutron flux distribution, power peaking factor, excess reactivity, core shutdown margin, safety reactivity factor and temperature reactivity coefficient have been calculated and analyzed. The effect of cladding material on thermal-hydraulic parameters such as the temperature distribution of different parts of the cell and coolant are also analyzed. In view of the results, SiC cladding can be introduced as a more appropriate choice for this reactor compared with other materials.

Keywords: Cladding, Neutronic Parameters, Thermal-hydraulic Parameters, Tehran Research Reactor

Topic: Advanced Fission Systems

18 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-54] MORPHOLOGY OF Y-TI NANO-OXIDES IN ODS ALLOYS IRRADIATED WITH HIGH ENERGY HEAVY IONS

J.H. OConnell (a*), V.A. Skuratov (b), A.S. Sohatsky (b), K. Kornieieva (b), A.D. Volkov (c), M. Zdorovets (d)

a) CHRTEM, NMU, Port Elizabeth, South Africa *[email protected] b) FLNR, JINR, Dubna, Russia c) JSC VNIINM, Moscow, Russia d) Institute of Nuclear Physics, Nur-Sultan, Kazakhstan

Abstract

Oxide dispersion strengthened ferritic martensitic steels (ODS) are considered as candidates for fuel claddings for Gen IV nuclear reactors. The radiation tolerance of ODS steels is considered to be due to trapping of lattice defects and helium atoms by oxide particles and fine grain boundaries. When used as fuel cladding, these materials will be in close proximity to fissile fuel and exposed to fission fragment irradiation. Recent experiments demonstrated that heavy ions of fission fragment energy may induce amorphous latent tracks in Y-Ti oxides and at present there is no experimental data demonstrating that amorphized nanoparticles in ODS materials will assure the same properties and the same excellent radiation resistance as observed for steels containing crystalline nanoparticles. A lot of data related to swift heavy ion induced changes in morphology and properties of metal and semiconductor nanoparticles in oxide matrices are known from the literature. However, almost nothing is known about oxide nanoparticles in metallic matrices irradiated with high energy heavy ions.

The aim of this report is to summarize recent experimental results on the morphology of swift (167 and 220 MeV) Xe ion induced latent tracks in Y2Ti2O7 nanoparticles within ODS alloys during post-irradiation heat treatment and after irradiation at different temperatures.

Keywords: ODS, fission fragments, cladding

Topic: Advanced Fission Systems

19 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-56] A THORIUM-FUEL PIN NEUTRONIC ANALYSIS USING DIFFERENT NUCLEAR CODES

Felipe M. G. Pereira, Renato V. A. Marques, Márcia S. Santos, Carlos. E. Velasquez and Claubia Pereira

Departamento de Engenharia Nuclear - Universidade Federal de Minas Gerais Av. Antonio Carlos, 6627 campus UFMG 31.270-901, Belo Horizonte, MG

Abstract

Several different nuclear codes have been used to perform depletion and criticality calculations, already widespread among worldwide researchers. The neutron transport and depletion codes have their particularities such as the number of energy groups and multigroup cross section data included for each code. Therefore, this work aims to validate the model and cross sections data generated at DEN/UFMG using NJOY99 adopting a thorium fuel pin benchmark performed by MIT, INEEL and Czech Technical University, and using different computational nuclear codes. The validation consists in comparing results from codes and reference using benchmark methodology in criticality and depletion situations. To perform criticality at steady state and depletion calculations are used MCNPX, MCNP5, SERPENT, SCALE6.0, and MONTEBURNS. Besides that, an extension of the benchmark calculations is performed and nuclear reactor safety parameters are calculated for developed model. In this work are evaluated quantities such as the effective delayed neutron fraction, fuel temperatures coefficients and production and transmutation rates for each code considering fresh fuel and depletion situations. It is achieved effective delayed neutron fractions that decreased responding to changes in fuel composition and infinite multiplication factors that began simulation with lower differences than the ones obtained at burnup end, both results are a reflection of production and transmutation rates considered by each code.

Keywords: Thorium;Nuclear codes;Validation;Criticality calculation;Cross sections data;Depletion;Infinite multiplication factor; Effective delayed neutron fraction

Topic: Advanced Fission Systems

20 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-59] OVERVIEW ABOUT THE STUDIES ON ADVANCED NUCLEAR REACTORS PERFORMED AT THE FEDERAL UNIVERSITY OF MINAS GERAIS

A. A. P. Macedo, M. Gilbert, A. L. Vieira, A. A. Cunha, G. H. P. Dias, M. C. Ramos, M. E. Scari, F. C. Silva, P. A. L. Reis, C. A. M. Silva, A. L. Costa, M. A. F. Veloso, C.E. Velasquez and C. Pereira

Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais Av. Antônio Carlos, 6627, Campus UFMG PAC 1 – Anexo Engenharia, Pampulha, 31270-901, Belo Horizonte, MG, Brasil

Abstract

There is growing understanding that nuclear power plants are needed to complement intermittent energy sources and to decrease the emissions of carbon dioxide. However, the next generation reactors will need to incorporate innovative solutions regarding the radioactive wastes, safety improvements, proliferation-resistance, sustainability, efficiency, and cost. In the last decades the nuclear power industry has been developing and improving reactor technology as the projects of the next generation of nuclear power reactors. The advanced reactors concepts offer significant potential benefits because they can provide reliable, safe, clean energy as part of a mix with current nuclear technology. In this scope, the Nuclear Engineering Department of the Universidade Federal de Minas Gerais - Brazil (DEN-UFMG) has studied advanced nuclear systems, in the last years. This paper presents the results of advanced nuclear reactors concepts as the Very High Temperature Reactor (VHTR), Molten Salt Reactor (MSR), Sodium Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR) and Advanced CANDU Reactor (ACR). These studies evaluate the nuclear system at steady state and during burnup using models developed at DEN-UFMG. Moreover, studies about the thermal hydraulic behavior of advanced nuclear systems as the HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) e LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor), developed at the DEN-UFMG, are also presented. The nuclear code systems such as SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0, MCNP5, RELAP5-3D, WIMS-D and others, have been used in the investigations, mainly related to the safe operation. In this way, the aim of this paper is to present an overview about the research activities developed by the Department of Nuclear Energy at the Universidade Federal de Minas Gerais on Advanced Nuclear Systems.

Keywords: advanced nuclear reactor; VHTR; MSR; SFR; GFR

Topic: Advanced Fission Systems

21 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-67] NEUTRONIC ANALYSIS OF SODIUM-COOLED FAST REACTOR (SFR) DESIGN WITH VARIOUS FUEL TYPES USING SHUFFLING STRATEGY Mohammad Ali Shafii(a*), Revina Septi (a), Feriska Handayani (a), Artoto Arkundato (b), Zaki Su’ud (c)

a) Department of Physics, Andalas University Padang Indonesia *[email protected] b) Department of Physics, Jember University, Jember Indonesia c) Nuclear and Biophysics Laboratory, Bandung Institute of Technology Bandung Indonesia

Abstract

Neutronic analysis of Sodium-Cooled Fast Reactor (SFR) design with variations of fuel types using radial fuel suffling strategy has been investigated. One type of generation IV reactor that currently being researched for commercial implementation is SFR. In this research, the SFR design utilizes natural uranium as a fuel input. The reactor core is designed in the form of two-dimensional cylindrical geometry for various type of fuel such as MOX, UN-PuN, and U-Zr. Radial suffling strategy in the direction of R- Z axis is applied to SFR to manage the fuel burn up for long life reactor with natural circulation as a fuel cycle input. The burn up process is follows the fuel region movement scheme. The designed reactor core is divided into 10 regions, representing the reactor is operated for 10 year without refueling, where the volume in each region is made equal to one another. At beginning, the first region of reactor core is filled with natural uranium fuel as a input and it is called by first fuel cycle. The scheme just needs natural uranium as a fuel cycle input every beginning of 10 year of cycle. Furthermore, the fuel movement scheme is carried out for several types of fuel. The global neutronic parameters such as multiplication factor (keff) and burn up analysis are observed and optimized. Overall, in the output power of 550 MWt, the results indicate that U-Zr is the most optimal fuel to be applied and a greater chance of being operated for SFR.

Keywords: SFR, fuel type, multiplication factor, burn up, shuffling strategy

Topic: Advanced Fission Systems

22 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-68] SIMULATION OF DIFFERENT DENSITY MIXING PEBBLE FLOW IN A TWO- DIMENSION CIRCULATING PACKED BED SYSTEM

Dwi Irwanto, Sparisoma Viridi

Department of Physics, Institut Teknologi Bandung, Bandung 40132, Indonesia

Abstract

Using molecular dynamics method simulation of pebble flow in a two-dimension circulating packed bed system, where the bed are mixture of three different densities, is reported in this work. Basis of the system is real pebble bed reactor HTR-10, which is simplified to a two-dimension simulation system from its real three-dimesion simulation sytem (Wu et al., 2019). Two types of force are considered in this work. The first is normal contact force in the form of linear spring-dashpot (Schaefer et al., 1996) and the second is earth gravitation force. Friction force is neglected for simplicty, where it can induce rotation of the spherical fuel elements. Each type of element with different density enter the system in different radial position. Variations of density area performed in order to observe their influence to element radial positiion as it circulated in to (at the top) and out from (at the bottom) the system. It is observed that there is a weak relation between element density and its stable radial position.

Keywords: molecular dynamics, HTGR, circulating bed, pebble flow

Topic: Advanced Fission Systems

23 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-72] THE CONCEPTUAL DESIGN OF THORIUM-BASED MOLTEN SALT ENERGY AMPLIFIER

Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*) a) Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China b) CAS Center for Innovation in Advanced Nuclear Energy, Shanghai 201800, China *[email protected] c) ShanghaiTech University, Shanghai 201210, China

Abstract

From the previous research, we know that thorium-uranium fuel cycle system, i.e. U233 for fissile fuel and Th232 for breeding material, is hardly to achieve self-sustaining state in the moderated molten salt reactor without the complex online post-processing system. In our research, we use a proton accelerator to drive a thorium-based fast neutron molten salt subcritical reactor that improves the neutron efficiency in the system. The research results show that the molten salt energy amplifier driven by the proton accelerator we designed can achieve a long-term stable state, more than 10 years, under a rated power and a stabilizing k value without any online post-processing system and online replenishment of fuel. A physical design of the most simplified single loop molten salt energy amplifier was accomplished. Through the burnup calculation, a rated power 300 MWth molten salt energy amplifier will continue to run for 30 years without any online processing but inputting a 1 GeV proton beam within 4 mA during the whole operation period. And the temperature coefficient of the molten salt reactor is totally negative in the whole period.

Keywords: energy amplifier, molten salt, self-sustaining, proton accelerator

Topic: Advanced Fission Systems

24 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-74] ANALYSIS OF INTEGRATED TARGET IN THE THORIUM-BASED MOLTEN SALT ENERGY AMPLIFIER

Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*)

a) Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China b) CAS Center for Innovation in Advanced Nuclear Energy, Shanghai 201800, China *[email protected] c) ShanghaiTech University, Shanghai 201210, China

Abstract

Conventional neutron targets in Accelerator Driven Sub-critical System (ADS) are based on the design using heavy metal like lead. Considering high density of energy deposition on the target, it generally needs a separated loop for the target cooling. In the thorium based molten salt energy amplifier, an integrated target is conceived. Based on liquid fuel properties of the core and neutron productivity of the molten salt fuel by bombarding with the proton, the proton beam is direct led in the molten salt core without a separated single loop for the neutron target. The analysis of the design is presented, including the neutron productivity calculation and the thermal hydraulics simulation. Besides, evaluation of different layouts of the proton beam introducer in the core is presented to compare the influence of the neutron utilization efficiency in the system, depend on which we can choose different types of the beam introducer, with or without a beam window.

Keywords: integrated target, energy amplifier, molten salt fuel, proton beam introducer

Topic: Advanced Fission Systems

25 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-75] INVESTIGATION OF THORIUM- ALTERNATIVE FUEL MIXTURES IN A GAS TURBINE MODULAR HELIUM REACTOR

Sumer ŞAHİN (a), Hacı Mehmet ŞAHİN (b*), Özgür Erol (c)

(a) Bahcesehir University, Departmen of Energy System Engineering, İstanbul, TURKİYE

(b) University of Karabük, Departmen of Energy System Engineering, Karabük, TURKİYE *[email protected]

(c) Baskent University, Departmen of Mechanical Engineering, Ankara, TURKİYE

Abstract

In this study, Gas Turbine - Modular Helium Reactor (GT-MHR), one of the new generation reactors has been investigated because it has many advantages. Very high efficiencies can be achieved due to the gas coolant, and additionally alternative fuels can be utilized in this reactor. It has also lower waste quantity and higher safety margins. Moreover, one of the most important characteristics that possibility of usage of weapons-grade plutonium (WGrPu), reactor grade plutonium (RGrPu) and minor actinide (MA) with fertile fuel types (natural uranium and thorium) is a good option to the efficient usage as an alternative fuel mixture in a GT-MHR.In this purpose, utilization of natural uranium (nat-U) and thorium as fertile fuels has been analyzed using alternative fuels (WGrPu, RGrPu and MA) as driver fuel. Then, possibility of utilization of the alternative fuels/fertile fuels mixture was investigated and an optimum mixture ratio was determined. Therefore, a neutronic analysis for the full core reactor was performed by using MCNP5 with ENDF/B-VI cross-section library. Different mixture ratios were tested in order to find the appropriate mixture ratio of fertile and fissile fuel particles that gives a comparable keff value of the reference uranium fuel. Time dependent calculations were performed by using MONTEBURN2.0 with ORIGEN2.2 for each selected mixture. Calculations showed that, a GT-MHR type reactor, which is using the original TRISO fuel particle mixture of 20% enriched uranium + natural uranium (original fuel) has an effective multiplication factor (keff) of 1.27. Corresponding to this keff value the alternative fuels/fertile fuels mixture was found as ratio of percent. In the final analysis, different parameters (operation time, burnup value, fissile isotope change, etc.) were subject of performance comparison.

Keywords: Thorium, Alternative Fuels, GT-MHR, MCNP

Topic: Advanced Fission Systems

26 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-91] MODIFIED CANDLE ANALSYSIS USING MICROSCOPIC CROSS SECTION FROM SLAROM CODE FOR DETAIL ANALYSIS OF MULTI SCENARIO MODIFIED CANDLE SCHEME

Zaki Su’ud, Fitria Miftasani, Feriska H. Irka, Nina Widiawati, Helen Raflis, H Sekimoto, Sumer Sahin, Mehmed Sahin, Zuhair

1Nuclear and Biophysics Research Divisions, Bandung Institute of Technology 2Emeritus Professor, Tokyo Institute of Technology, Japan 3Bachcisehir University, Turkey 4National Nuclear Energy Agency ( BATAN), Indonesia Email: [email protected]

Abstract

Implementation of Modified CANDLE burnup scheme based on microscopic cross section from SLAROM code has been implemented using two dan three dimensional analysis including iterative multi-group diffusion and burn-up analysis. In the previous calculation based on SRAC, the burn-up analysis generally performed for each region or sub region bases, while in this study the burn-up calculation is performed for each individual mesh. Therefore more flexible model of Modified CANDLE can be implemented including pure axial shuffling, pure radial shuffling, and also various combination of axial-radial shuffling can be implemented with or without assumption of special adjustment process in the pin level to minimized burn-up peaking. The general simulation scheme, initially microscopic cross section is generated by SLAROM code system to generate sets of microscopic crosss sections. Then multigroup diffusion calculation is performed and then continued by burnup analysis for every 10 years of period. After 10 years of burnup the fuel material in the core are shifted according to the detail model of Modified Burn-up scheme (axial shuffling, radial shuffling or combined axial-radial shuffling). Some calculation results show that in general for the same conditions, the calculation using this system agreed with those in the previous model. More detail results will be discussed in the conference including comparison of CANDLE, precious Modified CANCLE model and current Modified CANDLE models.

Keywords: Modified CANDLE, Microscopic cross section, iterative scheme, shuffling strategy

Topic: Advanced Fission Systems

27 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-92] PRELIMINARY STUDY OF NUCLEAR SAFETY SYSTEM ANALYSIS AND SIMULATION FOR MOLTEN SALT REACTOR

Muhammad Ilham (a), Cici Wulandari (a), Putranto Ilham Yazid (a), Sidik Permana (a)

a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, INDONESIA

Abstract

The development of technology and Molten Salt Reactor (MSR) research in the world has increased in the 2000s. For safety factor, MSR operates at low pressure which reducing the risk of pipe rupture or leak when an accident occurs. If the temperature increase over the design limit due to accident, there is an emergency cooling dump underneath the core as a safety where all the melting fuel will fall into the dump. MSR has large negative feedback reactivity when there is an expansion in fuel volume due to the temperature exceeding the design limits. In this study, point kinetic modeling is performed using reactor transient conditions where the fuel salt is circulating in core and loop. An analysis was conducted on the FUJI-12 type when there was a positive reactivity added in to evaluate the safety system. The calculation is verified and give a similar trend result with the previous study. The results obtained will serve as the minimum limits as the benchmark to design the MSR safety system.

Keywords: MSR, Point kinetic, safety system, simulation

Topic: Advanced Fission Systems

28 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-93] DESIGN DEVELOPMENT OF PELUIT-40 A SMALL COGENERATION NUCLEAR POWER PLANT

Topan Setiadipura(a*), Dwi Irwanto(b), Arya Adhyaksa Waskita(a), Hery Adrial(a), Suwoto(a), Zuhair(a)

(a)Centre for Nuclear Reactor Technology and Safety – National Nuclear Energy Agency, Puspiptek Area Building No. 80, Serpong, South Tangerang, Indonesia 15310 (b)2Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Indonesia, Jalan Ganesha 10 Bandung 40132, INDONESIA

Abstract

Economic performance is among important aspects for the success of a nuclear reactor design to be applied. A follow-up design based on Reaktor Daya Eksperimental (RDE) is developed in this study from the initial 10MWt power of RDE, current design will be upscaled to class of 30MWt without any update on the fuel and core geometry. Current design is called PeLUIt-30. This triplet power improvement hopely will increase the economic feasibility of the small modular pebble bed reactor. The important parmeter as the constrain of the design upscaling is the maximum temperature of the fuel in equilibrium condition also in the depressurized loss of forced-cooling (DLOFC) which assumed to be the severest accident hyptothetical scenario. PEBBED code is utilized for the equilibrium core analysis including the neutronic and thermal hydraulic module. In current study the material composition of the pebble fuel is maintained to assure the direct practical application of this design as the commercial follow-up of the RDE. Results of this study show that improving the power to 40MWt with a discharge burnup of 80 MWd/Kg-HM will have a maximum fuel temperature at equilibrium and DLOFC of 1037.4°C and 1562°C, respectively. These results show that PeLUIt-40 able to maintained the sound passive safety of the RDE while improving its commercial feasibility with higher power output.

Keywords: pebble bed reactor, high temperature gas-cooled reactor, cogeneration

Topic: Advanced Fission Systems

29 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-5] PRELIMINARY CALCULATION ON THE CONTAINMENT EXTERNAL COOLING EFFECT FOR FLEX STRATEGY USING CONTAINMENT ANALYSIS CODES

KYUNGHO NAM, Bum-Soo Youn

KHNP Central Research Insitute

Abstract

The external containment cooling strategy is involved in the FLEX Support Guideline (FSG). According to this document, the external containment cooling strategy will be most effective if the steel containmnet vessel itself can be sprayed with cool water. additionally, this cooling strategy should be evaluated for plant-specific containment building design. In case of Korean nuclear power plant, the material of containment building is pre-stressed concrete. Therefore, it should be checked that the external cooling strategy which is specified in FSG has an effect on the depressurazation of containment building. In this paper, the containment external cooling effect was anlyzed using GOTHIC and CAP code. In order to invertigate the influence of external cooling, an Extended Loss of All AC Power (ELAP) condition which is one of the entry condition for FSG-12 was applied. Additionally, the maximum RCP leakage was also assumed. As a calculation results, it was showed that the external spray cooling effect using portable pump have little effect on depressurization of containment building after 48 hours. And, the containment pressure is low sufficiently to maintain the containment integrity and implement other mitigation strategies.

Keywords: Containment, FSG, External Cooling, GOTHIC, CAP

Topic: Advanced Technology and Other Issues

30 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-6] THORIUM IN BRAZIL

Jose Rubens Maiorino

Federal University of ABC Abstract

Just after the discovery of thorium by Jöns Borzelius, a Sweden in 1829., the exploitation of thorium from monazite sand in Brazil date back to 1886, when Englishman John Gordon began exporting to Europe the ore mined in the municipality of Prado, Bahia State to use in lighting (incandescent gas lamps). In the late of 19th and early of 20th century, the interest in monazite increased owing to the use of thorium nitrate by gas mantle industries. Later, the use of lanthanide elements turned monazite into a much more important commodity than it was in pre-war years. The commercial exploitation of monazite sand starts in 1948, by a private company called ORQUINA in São Paulo city, and later the production and purification of thorium compounds was carried out at IPEN, a Research Institute in Sao Paulo, for about 18 years. The raw materials used were some thorium concentrates obtained from the industrialization of monazite sands, a process carried out in Sao Paulo between 1948 and 1994 on an industrial scale by the company ORQUIMA, later by a state company NUCLEMON (acronym for Nuclear Monazite), which operates up today.The first national program to use thorium was conducted during the 60’ by a research group from a Brazilian State, Minas Gerais, very rich in mineral resources, including thorium. This research group was called the “Thorium Groupâ€, and in the framework of a cooperation agreement with the French CEA aimed at the development of a thorium fueled PHWR with a concept of a pre stressed concrete reactor vessel. Also, in the beginning of the seventies, in the frame work of a cooperation agreement of IPEN, in São Paulo, with the USA General Atomic (GA), several activities, theoretical and experimental, were developed on thorium technology and utilization mainly for the HTGR concept. However, it was in the framework of the Brazilian German agreement that the biggest R&D program on thorium utilization was developed with the incentive of “International Nuclear Fuel Cycle Evaluationâ€. This program was conducted by the Brazilian Center for Development of Nuclear Technology (CDTN), in that time the R&D branch of the former holding, NUCLEBRAS, and the Germans KFA- Jûlich, Siemens A.G-KWU, and NUKEN.). This paper, besides the historical overview, will discuss the natural resources of thorium in Brazil, the technological capability to produce nuclear thorium, and the main results obtained in the previous national programs, as well as, the academic researchers on going at the Brazilian Universities. As main conclusion, given the huge reserves available in Brazil, the government should support researchers on thorium to be on line with this important energetic in the international community.

Keywords: Thorium, Monazite, Nuclear Reactors, Brazil Topic: Advanced Technology and Other Issues 31 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-9] THE CALCULATION OF THE NEUTRON YIELD DISTRIBUTION FOR THE GAS TARGET OF HINEG

Z. Yang (a, b), Y. Zhang (a), Z. Wang (a), S. Chen (a*), W. Wang (a), T. Li (a), Y. Wu (a), FDS Team

a) Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China *[email protected] b) University of Science Technology of China, Hefei, Anhui, 230031, China

Abstract

Intensive fusion neutron source have great value in the research of fusion nuclear energy system and the application of advanced nuclear technology. Using gas target instead of solid target is an important way to remarkably increase the yield of fusion neutron source. Based on the High Intensity D-T Fusion Neutron Generator (HINEG, built and developed by INEST, FDS Team), a windowless gas target with a 5 cm × 1 m reaction chamber has been designed and constructing. TheΦ accelerated deuterium ions beam enters the gas target along the central axis and travel through the reaction chamber. The neutrons release from the beam path in the target. In order to accurately estimate and optimize fusion neutron yield in gas target, Monte Carlo simulations have been done to calculate neutron yield and release spatial distribution with 200-400 keV incident deuterium ion beam energies and 800-1200 Pa target chamber pressures of deuterium and tritium separately. The work can provide reference parameters for the subsequent commissioning of gas target in HINEG, as well as the neutron release spatial distribution model for future neutron yield measurement experiment.

Keywords: HINEG, neutron yield, gas target, Monte Carlo simulation

Topic: Advanced Technology and Other Issues

32 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-11] A COUPLING CONTROL STRATEGY OF ACCELERATOR AND GAS TARGET FOR HIGH INTENSITY D-T FUSION NEUTRON GENERATOR

C. Zhao(a,b), J. Wang(a), Y. Wang(a), Z. Wang(a), Q. Zhang(a), Y. Zhang(a*), FDS Team(a)

a) Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China *yong.zhang2 @fds.org.cn b) University of Science and Technology of China, Hefei, Anhui, 230031, China

Abstract

High Intensity D-T Fusion Neutron Generator (HINEG), developed by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS), is an important experimental platform for the development of fusion technology. The challenge of D-T fusion neutron generator mainly focus on the high heat dissipation of the tritium target. Therefore, FDS Team developed a gas target to tackle the challenge mentioned above. The gas target can effectively improve the density of tritium particles, enhance the heat dissipation capacity, and can be operated continuously. It avoids changing the target in the operation process of D-T fusion neutron generator and improves the efficiency and economy. There are mainly two factors affecting the neutron yield, the incident deuterium ion current and the tritium particle density in the gas target. At present, the neutron yield remains constant depends on manual regulation of these two factors. In order to realize automation and intelligence regulation for the control for the neutron yield, a coupling control strategy for both the incident deuterium ion current and the tritium particle density in the gas target is proposed. The strategy can independently implement an optimal solution based on the set values of neutron yield through coordinated control of incident beam intensity and particle density of the gas target. The validity of the coupling control strategy has been proved by a large amount of experimental data. This coupling control strategy is helpful for the exact control of the neutron yield of D-T fusion neutron generator with the gas target. It can also improve the control precision and stability of the D-T fusion neutron generator.

Keywords: HINEG, Coupling control, Accelerator, Gas target, Control precision

Topic: Advanced Technology and Other Issues

33 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-21] THE AP-TH 1000 AN ADVANCED CONCEPT TO USE MOX OF THORIUM IN A CLOSED FUEL CYCLE

Giovanni Laranjo de Stefani; Jose Rubens Maiorino

Instituto de Pesquisas Energéticas e Nucleares (IPEN / CNEN - SP); Federal University of ABC

Abstract

This work presents a study for the firsts 4 cycles of recharge of the reactor AP-Th1000, a version of the reactor AP1000 using mixed uranium and thorium oxides as fuel, which the feasibility studies had been already demonstrated in previous study for a first cycle. The AP-Th1000 study is a proposal to start the thorium fuel cycle using the most common reactor technology in the nuclear industry, the Pressurized Water Reactors (PWR). A closed cycle study is carried out for the first 4 cycles where sustainability parameters such as the use of natural resources, reduction of long-lived actinides and production of 233U are evaluated. In cycles 2, 3 and 4, new assemblies with a fuel of the remaining uranium from the previous cycle are used instead of assemblies removed from the core, thus being a mixture of different uranium’s (232U, 233U, 234U, 235U, 236U and 238U) , where the additional fissile material inserted into the fuel to ensure the 18-month operation of the reactor comes from uranium oxides enriched at 20 w / o. Neutron parameters such as the fraction of delayed neutrons (βeff), temporal evolution of the effective multiplication factor (keff) and temperature reactivity coefficient are evaluated to guarantee the feasibility of operation of the proposed reactor. The results demonstrate the viability of the proposal and a gain in sustainability using closed fuel cycle.

Keywords: Thorium, PWR, Closed Fuel Cycle, AP 1000

Topic: Advanced Technology and Other Issues

34 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-22] METALLIC FISSION PRODUCT TRANSPORT IN TRISO COATED PARTICLES

Johannes Neethling (a*), Jaco Olivier (a), Jacques OConnell (a)

a) Centre for HRTEM, Department of Physics, Nelson Mandela University, Port Elizabeth, 6001, South Africa *[email protected]

Abstract

The characterization of tristructural-isotropic (TRISO) coated fuel particles using transmission electron microscopy has been successfully used in South Africa since 2006. TRISO coated fuel particles are used in high temperature gas cooled nuclear reactors (HTGRs). The polycrystalline 3C-SiC layer in the TRISO particle acts as the main barrier to fission product release. The finding, more than three decades years ago, that silver (a radioactive fission product) can be released by reputedly intact TRISO nuclear fuel particles has led to significant research efforts to determine the silver transport mechanism in 3C-SiC. Neethling and co-workers discovered that palladium, another high yield fission product, significantly enhances the transport of silver along grain boundaries in SiC [1]. The Pd assisted Ag transport mechanism was confirmed in a recent paper by Van Rooyen, Olivier and Neethling [2]. In this paper analytical and high resolution STEM was used to examine the microstructure and location of fission products (Pd and Ag) in 3C-SiC obtained from a neutron irradiated TRISO particle. It was found that Pd formed a palladium silicide layer along the SiC grain boundaries. It was suggested that palladium penetrates the SiC along grain boundaries and reacts with the SiC to form palladium silicide, which creates fast diffusion paths for Ag. This mechanism together with the measured diffusion coefficient of Ag in palladium silicide, agrees closely with the Ag release rates obtained from irradiated and annealed TRISO particles during the past 4 decades.In this talk the highlights of our research on TRISO particles and metallic fission product transport will be presented. A brief comparison of the technical challenges of He cooled pebble bed reactors and molten salt reactors will also be presented.

References [1] JH Neethling, JH O’Connell and EJ Olivier, Nucl. Eng. and Design 251 (2012), p. 230. [2] IJ Van Rooyen, EJ Olivier and JH Neethling, J. Nucl. Mater. 476 (2016), p. 93.

Keywords: TRISO coated particles, transmission electron microscopy, fission product transport in SiC

Topic: Advanced Technology and Other Issues

35 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-32] THERMAL BEHAVIOR STUDY OF VERTICAL HEATER VERTICAL COOLER (VHVC) NATURAL CIRCULATION LOOP (NCL) WITH HORIZONTAL WIDTH VARIATION

(1)(a)Duwi Hariyanto and (1)(2)(b)Sidik Permana

(1) Nuclear Physics and Biophysics Research Division, Institut Teknologi Bandung (2) Nuclear Science and Engineering Research Division, Institut Teknologi Bandung Email : (a) [email protected], (b) [email protected]

Abstract

Natural circulation loop is one of the design concepts of cooling system in new advanced reactors that has attracted many researchers to develop it. The aim of this study is to perceive the effect of horizontal width variation on the thermal behavior of a single- phase natural circulation loop (NCL). COMSOL multyphysics software has been used for numerical study and a NCL apparatus with vertical heater and vertical cooler has been designed for experimental study. The heater has been designed using nicrome wire on the outside of the stainless pipe. The cooler has been arranged using pipe-in-pipe with water flowing through the annulus. Arduino microcontroller and K-type thermocouple sensors have been used in temperature data acquisition. The initial numerical and experimental results show a stable thermal behavior along the loop. This study is supposed to be one of references for single-phase natural circulation loop with vertical heater and vertical cooler.

Keywords: single-phase, natural circulation, thermal behavior

Topic: Advanced Technology and Other Issues

36 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-33] DYNAMIC VISUALIZATION ON EROSION BEHAVIOR OF A SOLID PLATE WITH PARTICLE METHOD

M Ifthacharo(a*), A P A Mustari(b), S Permana(b), A Nuril(a)

a) Master Program of Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia *[email protected] b) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia

Abstract

There are many systems in a reactor shutdown function in MSR in addition to inherent self-stabilization. One of those systems is the fuel-salt drain system. The present study focused on the melting and solidification phenomenon that occurs in the freeze valve. An experiment was performed to investigate the erosion behavior of a solid plate by an impinging liquid with respect to time. In addition, a numerical modelling based on MPS method to visualize the occurring erosion will also be carried out. The experiment will be conducted by varying the parameters such as the liquids, temperature, and diameter. Hot water (70-80oC), molten paraffin, and molten candle wax, will be used while both molded candle and paraffin will serve as the target plates. The dimension of the target plate is a cylindrical with 44 mm in thickness and 140 mm in length for both paraffin and candle wax. From the simulation, the distribution and heat transfer of the erosion along with the pool effect formation will be studied and analyze. Time of the erosion of the experiment will be compared to the erosion time acquired in the simulation.

Keywords: MPS; erosion behavior; Freeze-valve

Topic: Advanced Technology and Other Issues

37 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-38] DEVELOPMENT OF A NEW HYBRID METHOD TO ASSESS THE FREQUENCY OF OCCURRENCE OF LOSS OF OFFSITE POWER TO THE NUCLEAR POWER PLANTS

Shahabeddin Kamyab 1; Mohammadreza Nematollahi 1,2; Faramarz Yousefpour 3

1. School of Engineering, Shiraz University 71348-51154, Shiraz, Iran 2. Safety Research Center of Shiraz University 71348-51154, Shiraz, Iran 3. Nuclear Science & Technology Research Center, Tehran, Iran

Abstract

The unavailability of offsite electrical power, grid, to the nuclear power plant (NPP) is a major challenge to their safety. In this regards, Loss of Offsite Power (LOOP) is concerned as one of the most important undesired events from the safety analysis perspective. For this, probabilistic safety assessment (PSA) demands for the occurrence frequency of LOOP as an initiating event, in order to assess its contribution in total core damage frequency (CDF) of a NPP. The mutual dependence between the availability of the offsite power and safe operation of nuclear power plant makes the issue much more intricate. Over the decades, plenty of studies have surveyed the reliability evaluation issue of the grid, which culminated in proposal of various outputs and indices. This article is, first, devoted to an introductory review over the available methods. It compares the benefits and drawbacks of each, and concludes with insufficiency and impotency of them in satisfying the PSA demands. Thereafter, a new hybrid method is developed to estimate the frequency of occurrence of LOOP as a relevant input into the PSA model. The method combines the deterministic results of dynamic stability analysis to confirm the probabilistic outcomes of the PSA model of the grid. Relying on the new method helps to identify the risk-significant deficiencies of the offsite electrical connections to the NPP, to prioritize them based on the result of importance measures and exploiting risk-informed decision making advantageous to modify the operation and design of the connected grid.

Keywords: Loss of Offsite Power; Probabilistic Safety Assessment; Stability Analysis, Grid Configuration; Risk Informed Decision Making

Topic: Advanced Technology and Other Issues

38 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-39] NUMERICAL AND 2D PIV STUDY OF THE EFFECTS OF HYDRODYNAMIC PARAMETERS ON THE FLOW ACCELERATED CORROSION DOWNSTREAM OF A GATE VALVE

Abbas Sedghkerdar1, Mohammadreza Nematollahi 1, 2 * and Ali Erfaninia 2, 3

1 Shiraz University, School of Mechanical Engineering, Mollasadra St, 71348-51154, Shiraz, Iran. 2Safety Research Center, Shiraz University, Mollasadra St, 71348-51154, Shiraz, Iran. 3Persian Gulf University, Center for Nuclear Energy Research, Mahini St, Bushehr, Iran.

Abstract

A lot of disastrous failures have been reported at several nuclear power plants around the world since 1981 due to FAC. The convective mass transfer of the iron spices in the water is the main phase of Flow Accelerated Corrosion (FAC). Gate valve is a component that is extremely prone to FAC. In this study the flow field downstream of a typical gate valve is visualized by developing a test facility and using 2D PIV. The effect of flow velocity on the mass transfer coefficient downstream of a typical gate valve is investigated via CFX simulation after validating the CFX by 2D PIV results. The SST model shows a very good ability to flow field simulation downstream the gate valve. The effect of the flow velocity on the mass transfer coefficient is studied and the all locus of the local mass transfer coefficient peaks are determined. The results is very important for maintenance programs and reliability evaluation of NPPs piping system.

Keywords: Keywords: Mass Transfer Coefficient, FAC, Gate Valve, 2D PIV, CFX

Topic: Advanced Technology and Other Issues

39 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-40] GAMMA HEATING EVALUATION OF RSG G.A. SIWABESSY SILICIDE CORE

Anis Rohanda1,3, Abdul Waris2, Rizal Kurniadi2, Syaiful Bakhri3

1 Department of Physics, Institut Teknologi Bandung Jl. Ganesa No. 10, Bandung 40312 2Nuclear Physics and Biophysics Research Division,Department of Physics, Institut Teknologi Bandung Jl. Ganesa No. 10, Bandung 40312 3Center for Nuclear Reactor Technology and Safety (PTKRN) National Nuclear Energy Agency of Indonesia (BATAN) Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 E-mail: [email protected]

Abstract

Reaktor Serba Guna G.A. Siwabessy (RSG-GAS, previousname MPR-30) is a research reactor that serves as a place to irradiate various type of material. In each material irradiation activity gamma heat analysis is needed as part of material target safety analysis of the irradiation facility. RSG-GAS is designed to have a 30 MWt nominal power but it is currently operated at a 15 MW power level. Conversion of RSG-GAS core from oxide fuel elements to silicides has been carried out. The purpose of the conversion are to improve the performance and efficiency of the RSG-GAS. In this study, the calculation of RSG-GAS silicides core gamma heat was conducted in 15 MW and 30 MW power level at E7 Central Irradiation Position (CIP). The calculation was performed by using Gamset code to verify the measurement results based on gamma calorimeter method on 4 types of absorbent materials, namely graphite (C), aluminum (Al), iron (Fe) and zirconium (Zr). The results of this verification act as preliminary data for the development of a gamma heating calculation program. Evaluation of the results shows that calculated gamma heat has an upward trend which corresponds to the measured gamma heat, along with the increase in the target material atomic number and reactor power. The results show the smallest difference of 1.42% in the Al target material at 30 MW. In general, the measurements results are lower than the calculation results. This corresponds to the results of gamma heat benchmarking on the RSG-GAS oxide core.

Keywords: RSG-GAS, gamma heat, Gamset, silicide core

Topic: Advanced Technology and Other Issues

40 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-41] THE MOLTEN WOODS METAL INITIAL VELOCITY VARIATIONS EFFECT ON BREACHING PROCESS

A N Hidayati (*a), A Waris (b), A P A Mustari (b), N A Aprianti (b), M Iftacharo (c)

a) Doctoral Program of Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia *[email protected]

b) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia

c) Master Program of Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia

Abstract

Series of MPS simulations have been conducted using two dimensional geometry. The simulation was based on Sudha’s experiment (2018) about initial velocity variations on molten Wood’s Metal (WM). The molten WM would be flowed through nozzle with the diameter was 6 mm. It would impinge to the Woods Metal Plate (WMP) which 270 mm below the nozzle. The WMP diameter was 470 mm. The temperature of molten WM and WMP were set at 573 K and 300 K, respectively. The initial velocity of molten WM was varied at 0,327 m/s, 0,397 m/s, 0,498 m/s in the y-negative direction. The simulation was calculated by using 2D MPS with additional procedures such as heat transfer calculation and defining a new type of wall particle. The results showed some different spread patterns, leading edge and phase fraction change for each initial velocity.

Keywords: velocity, descend, MPS, density, temperature

Topic: Advanced Technology and Other Issues

41 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-44] OPPORTUNITIES OF THE NUCLEAR ENERGY INTEGRATED WITH VARIABLE RENEWABLES

Emilio MINGUEZ

Institute of Nuclear Fusion Universidad Politécnica de Madrid

Abstract

The energy roadmap of the European Union proposes decarbonization the power sector by 2050, by means of the use of renewable energies, and avoiding those resources producing carbon dioxide. In some countries of EU the nuclear energy will be fully avoided due to generate electricity only with renewable energies.However, at the same time the main objective of the energy roadmap of EU is to have an energy mix sustainable from the environmental point of view, economical and with warranty of supply. In this sense our proposal is to combine small nuclear systems with variable renewable to generate electricity and also to provide thermal energy for many industrial applications.We propose a hybrid system composed by a Small Modular Reactors (SMR) or micro-reactors connected with a solar or/and wind energies. SMRs have a large number of benefits: passive safety, reduced financing they need because a quicker construction time, flexibility to add new modules and independent operation. Different options of connections with available renewables will be explained as a resume of the available proposed designs.SMRs could be an alternative option to substitute a coal fire power plant, which will give a continuous supply of power carbon-free. Then, nuclear energy in combination with renewables can provide not only electrical energy but also thermal power for other applications without carbon emissions. Available solutions and reductions of carbon emissions will be analyzed for a country like Spain in which in 2050 all electrical energy will be generated by renewables only according to political decissions. Expectations about development of devices for energy storage will be also analyzed.

Keywords: SMR, carbon free, decontamination

Topic: Advanced Technology and Other Issues

42 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-51] SIMULATION OF TRIGA 2000 BANDUNG NEUTRON AND GAMMA DATA DOSE USING MCNPX, PHITS AND EXTRAPOLATION WITH TENTH VALUE LAYER (TVL )

Rakotovao Lovanantenaina Oméga(a), Sidik Permana(b), Rini Heroe Oetami(c), Rasito(d)

(a) Master Program of Physics Department, Faculty of Mathematics and natural science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia (b) Nuclear Science and Engineering Dept., Physics Dept., Bandung Institute of Technology (c) PSTNT BATAN Bandung (d) PSTNT BATAN Bandung

Abstract

All use of a radiation sources should be monitored to prevent harmful effects of radiation on human health and contribute to the protection of the environment. TRIGA 2000 (Training, Research, Isotopes, General Atomics) Bandung is one of the research reactors own by BATAN which is one the division using radiation sources. The assessment of gamma and neutron doses is a part of the radiation protection policies in BATAN Bandung to optimize the radiation protection and avoid the risk that can affect the Worker, publics and Environment against the exposure. The purpose of this research is to calculate the dose from the core of the research reactor to the whole component of reactor with the consideration of all types of material used in the building. The calculation starts from modelling the geometry with MCNPX. Next is to extract the MCNPX output calculation into PHITS to get the variation of the dose from the core in such a distance and will be continued using Tenth Value Layer to assume the rest of the distance that should be calculated during the running of the program. The data given by the PHITS graph is plotted to excel to estimate the neutron and gamma dose for the whole TRIGA 2000.

Keywords: Gamma dose, neutron dose. MCNPX, PHITS, TVL

Topic: Advanced Technology and Other Issues

43 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-55] PRELIMINARY STUDY OF THE RADIONUCLIDES TRANSPORT BEHAVIOUR DURING NORMAL OPERATION OF DIRECT-CYCLE GAS TURBINE HTR

I Wayan Ngarayana 1,2; Kenta Murakami 1

1. Nuclear Material & Maintenology Labolatory, Nuclear System Safety Engineering, Nagaoka University of technology 2. National Nuclear Energy Agency of Indonesia

Abstract

The inherent safety feature of High Temperature Gas Reactor (HTGR) is not only making the core meltdown accident become almost impossible but also able to reduce the radioactive release probability both on normal and accident condition become relatively low. Nevertheless, radioactivity due to the release of the fission products including activated impurity nuclides leaves several important issues that need to be solved especially for direct-cycle Gas Turbine High Temperature Reactor (GTHTR) since movable and frequently access maintenance activities related component are located on the primary cooling system. Although many studies have been devoted to improving the understanding of the radionuclides transport behaviour of HTGR, there are still open areas that require in-depth understanding to reduce transport modelling uncertainty including the radionuclides distribution characteristics. We are identifying several issues from the current status of typical HTGR radionuclides transport studies including specific issues for GTHTR. From those issues, the future study direction as part of the maintenance strategy preparation of the GTHTR study is prepared.

Keywords: HTGR, Direct-cycle, Radionuclide, Transport, Behaviour, Maintenance

Topic: Advanced Technology and Other Issues

44 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-60] GRADING FOR THE MAINTENANCE ACTIVITIES OF THE ADVANCED NUCLEAR REACTOR USING MODIFIED FUZZY FMEA & EXPERT JUDGEMENT METHODOLOGY

I Wayan Ngarayana 1,2; Kenta Murakami1

1. Nuclear Material and Maintenology Labolatory, Nuclear System Safety Engineering, Nagaoka University of Technology 2. National Nuclear Energy Agency of Indonesia

Abstract

Grading is an important step of the Nuclear Power Plant (NPP) operation & maintenance activities. However, there are several grading difficulties for the advanced Generation IV NPP causing by the lack of operational experiences and availability of the reliability data. Failure Mode & Effects Analysis (FMEA) is one of the mature techniques that commonly used to solve such kind of difficulties. Nevertheless, traditional FMEA has several issues and possibly become an obstacle in the grading process. The modified FMEA by utilising expert judgement elicitation techniques combined with the fuzzy logic theory is proposed to solve those issues. As a study practice, the proposed methodology is applied by examining Japanese’s HTGR, Gas Turbine High Temperature Reactor 300 for Co- generation (GTHTR300C) design carefully. This study establishing good practice especially for the future advance NPP maintenance activities development.

Keywords: Advancec NPP, Maintenance, Grading, Fuzzy, Expert, Judgement, GTHTR300C

Topic: Advanced Technology and Other Issues

45 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-66] NEUTRONIC ASSESMENT OF HTGR STUDY DESIGN ON THE USE OF ZRC TRISO-COATED PARTICLE (TRIZO)

Fitria Miftasani, Zaki Suud, Dwi Irwanto

Bandung Institute of Technology

Abstract

The neutronic assesment has been studied to high temperature gas-cooled reactor that uses ZrC Triso-coated particle (TRIZO) in previous work. In the present study, the TRIZO is applied on the design parameter of HTGR with various power from 50 MWt-300 MWt. The ZrC coating layer is a strong candidate to replace the SiC coating layer of TRISO coated fuel particles to improve its endurance against irradiation and high temperature. The appropriate size of reactor core will be evaluated for reactor with powers of 50 MWt, 100 MWt, 150 MWt, 200 MWt and 300 MWt. Consequently, neutronic analyses for the proposed reactor core should be conducted. The neutronic aspect of the reactor is assessed by investigating the k-eff and k-inf . The materials used in this calculation was based on the standard TRISO fuel particle as being used in the HTTR 30 MWt reactor, with uranium dioxide fuels based.

Keywords: HTGR, ZrC, Power, Neutronic

Topic: Advanced Technology and Other Issues

46 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-69] E WALL EFFECT AND FAST NEUTRON IRRADIATION IMPACT IN THERMAL HYDRAULICS ANALYSIS OF HTR-10

Bilal El Bari(a*), Dwi Irwanto(a)

a)Nuclear Physics and Biophysics Research Group, Department of Physics, Institut Teknologi Bandung, Jl. Ganesha No. 10, Bandung, 40132, Indonesia *[email protected]

Abstract

Thermal hydraulics aspect is one of the crucial aspects that must be considered when reactor design and operation analysis were performed because this aspect involves security, safety, and efficiency factor that must have to be examined. In this study, thermal-hydraulics aspect of the HTR-10 reactor was analyzed by reviewing the effects of wall and fast neutron irradiation by modifying the PEBBLE code that using finite- difference numerical method for solving the differential equation of the system. In reviewing wall effect, the different porosity distribution alongside the edge was determined using Mueller’s model. As for the fast neutron irradiation effect, fast neutron doses that affect the properties of graphite material are taking into account in the calculation.

Keywords: Fast Neutron Irradiation; Finite-Difference; HTR-10; Thermal Hydraulics; Wall Effect

Topic: Advanced Technology and Other Issues

47 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-70] CHARACTERISTICS OF AUSTENITE SS 316L STEEL UNDER HIGH TEMPERATURE MOLTEN LEAD BISMUTH ENVIRONMENT: THE MOLECULAR DYNAMICS SIMULATION OF CORROSION INHIBITION

Artoto Arkundato(1), Fiber Monado(2), Mohammad Ali Shafii(3), Ratna Dewi Syarifah(4)

(1) Department of Physics, Faculty of Mathematical and Natural Sciences, University of Jember

(2) Department of Physics, Faculty of Mathematical and Natural Sciences, University of Sriwijaya (3)Department of Physics, Faculty of Mathematical and Natural Sciences, Andalas University (4)Department of Physics, Faculty of Mathematical and Natural Sciences, University of Jember

Abstract

The performance of Austenite SS 316L Steels with Cr and Ni content under high temperature molten lead bismuth coolant were investigated. The SS 316L with various Cr and Ni contents, and also the various composition of Pb-Bi eutectic collant was prepared. The interaction between the austenite steels and Pb-Bi eutectics were observed by molecular dynamics simulation. The corrosion of steel then is investigated and to avoid the high dissolution of steel components so an inhibitor (oxygen or nitrogen) must be injected into the coolant. The goal of the research is to know the characteristic of the SS 316L steel under various condition of temperatures, compositions of Pb-Bi coolant and also effect of concentration of inhibitor. The molecular dynamics simulation then can be used as a tool for prediction of the material properties under extreme condition where the experimental way is difficult and/or impossible. To support this research various softwares will be used as moldy, lammps, packmol, atomsk and ovito.

Keywords: Austenite steel, liquid metal corrosion, Pb-Bi eutectic, molecular dynamics

Topic: Advanced Technology and Other Issues

48 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-76] AN OVERVIEW OF THE APPLICABILITY OF SNI IEC 61331-1:2016 ON RADIATION APRON FOR MEDICAL RADIATION USE

Suzie Darmawati, Sunarto, Hanna Yasmine, Sigit Santosa

Center for Nuclear Standardization and Quality, National Nuclear Energy Agency of Indonesia

Abstract

The use of apron for radiation protection is regulated under the Indonesia Nuclear Regulatory Agency (BAPETEN) Decree no. 8 year 2011 about Radiation Safety and the Use of Diagnostic and Interventional Radiological X-Ray Machine. It listed the apron specifications are as follows: having thickness equivalent to 0.2 mm Pb or 0.25 mm Pb for diagnostic use and equivalent to 0.35 mm Pb or 0.5 mm Pb for interventional use. Further, National Standardization Agency (BSN) had issued SNI IEC 61331-1:2016, providing guidance for testing the plate materials on the apron using 400 kV x-ray machine and 1.3 MeV gamma exposure with narrow beam, to measure the attenuation ratio and air kerma rate The method used is to determine the attenuation ratio, build-up factors, and equivalent attenuation coefficient. There were 4 different aprons (A, B, C, and D) with 9 measurement points. The results showed the air kerma rate without apron was 0.664 mGy/second, the air kerma rate with lead-equivalent layer was 0.0006 mGy/second, and the best result was produced using the apron D, with the attenuation ratio ranging from 17.2 to 29.1, showing the most homogeneity.

Keywords: radiation apron; attenuation ratio; IEC 61331-1

Topic: Advanced Technology and Other Issues

49 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-79] CORROSION INVESTIGATION OF SILICON CARBIDE AND ZIRCONIUM CARBIDE CAUSED BY SILVER AT HIGH TEMPERATURE: A PRELIMINARY STUDY

Abu Khalid Rivai, Mardiyanto, Sumaryo, Bambang Sugeng, Nanda Shabrina

Center For Science and Technology of Advanced Materials National Nuclear Energy Agency of Indonesia - BATAN

Abstract

TRISO (Tri-Structural Isotropic) which consist of Inner Pyrolitic Carbon, SiC (Silicon Carbide) dan Outer Pyrolitic Carbon is one of the superiority of High Temperature Gas- cooled Reactor (HTGR) from safety systems feature viewpoint. However, one of the issues of the system is corrosion of SiC caused by silver (Ag). Zirconium Carbide (ZrC) as a extremely hard refractory ceramic material is a potential candidate to overcome the issue. In this study, the corrosion mechanism of SiC was investigated experimentally compared with ZrC. SiC and ZrC samples were tested in molten silver at high temperature more than 1000°C. The tested samples were analyzed using Microscope Optic (OM), Scanning Electron Microscope – Energy Dispersive X-ray Spectroscope (SEM-EDS) and XRD (X-Ray Diffractometer) to observe the corrosion mechanism.

Keywords: Corrosion, HTGR, TRISO, SiC, ZrC, Silver

Topic: Advanced Technology and Other Issues

50 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-80] EXPERIMENTAL INVESTIGATION ON CORROSION RESISTANT OF ZIRCONIUM CARBIDE COATING AT ELEVATED TEMPERATURE

Abu Khalid Rivai, Mardiyanto, Sumaryo, Bambang Sugeng, Nanda Shabrina

Center For Science and Technology of Advanced Materials National Nuclear Energy Agency of Indonesia - BATAN

Abstract

Zirconium Carbide (ZrC) as a refractory ceramic material has a superiority to be a coating materials of TRISO (Tri-Structural Isotropic) fuel system of High Temperature Gas-cooled Reactor (HTGR). The purpose is to replace of SiC (Silicon Carbide) material due to lack of corrosion resistance caused by the interaction with silver (Ag) and palladium (Pd) as fission products. In this study, ZrC has been deposited on a substrate made of steel using Plasma-Pulsed Laser Deposition (PLD) at Center For Science and Technology of Advanced Materials laboratory – National Nuclear Energy Agency of Indonesia (BATAN). ZrC was deposited with the chamber pressure around 235 mTorr, the substrate temperature of 850°C, the number of laser shots of 90,000 and the oxygen background gas with 40 sccm (standard cubic centimeters per minute). The ZrC-coated steel then tested at elevated temperature together with steel without coating for comparison analysis. Afterward, the before and after corrosion test samples were analyzed using Microscope Optic (OM), Scanning Electron Microscope – Energy Dispersive X-ray Spectroscope (SEM-EDS), XRD (X-Ray Diffractometer) and Atomic Force Microscope (AFM). The results showed that the ceramic coating could homogeneously and sticky deposited on the surface of the steel substrate surface after coating process. The change of the surface of ZrC-coated steel was different than the surface of the steel without coating after corrosion test.

Keywords: Corrosion, HTGR, ZrC, PLD, Coating

Topic: Advanced Technology and Other Issues

51 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-84] EVALUATION ON ENVIRONMENT NUCLEAR RADIATION AT VARIOUS CITIES IN JAVA ISLAND

Imam Ghazali Yasmint, Ahmad Lathiiful Quluub and Sidik Permana

Nuclear Physics Laboratory, Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha No. 10 Bandung, Indonesia, 40132

Abstract

In general, humans are exposed by various ionized radiation sources in their environment. Natural radiation sources are one of the greatest contributions to that exposure. Natural radiation comes from cosmic, terrestrial and internal radiations. In addition, some artificial radiations that come from the activities of nuclear reactor, hospitals and several industries have been also contributed. Java Island is the most populated area in Indonesia. Evaluation on natural background of nuclear radiation is important to be used as a basic data, which can be used to understand the radiation level and some source of contamination. Then, it can also produce a map of natural nuclear radiation in Java Island as information to the public about radiation doses. In this study, measurements were taken in various locations in Java Island, which popular and many visitor comes and do their activities. The evaluation is done by using GMC-320 Plus detector, which is based on a Geiger Muller detector. The result will be presented as a mapping area and the level of radiation in CPM and converted as radiation dose rate, which is shown as a normal of natural background radiation level.

Keywords: CPM and Dose Rate, GMC-320 Plus Survey Meter, Java Island, Natural Radiation, Radiation Map

Topic: Advanced Technology and Other Issues

52 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-85] CONSIDERATION OF INTRUSION STRATEGIES AND INSIDER-OUTSIDER COLLUSION IN ANALYZING PHYSICAL PROTECTION SYSTEM (PPS) USING A STOCHASTIC EASI MODEL

Yanuar Ady Setiawan

Department of Nuclear Engineering and Engineering Physics, Faculty of Engineering, Universitas Gadjah Mada

Abstract

The Estimate of Adversary Sequence Interruption (EASI) model is a single path analysis model to calculate the Probability of Interruption (PI) of the Physical Protection System (PPS) performance in securing a nuclear facility from theft, sabotage, or another malevolent adversary attack. Analyst utilizes Critical Detection Point (CDP) concept in analyzing the Adversary Sequence Diagram (ASD) to get the Most Vulnerable Path (MVP), an intrusion path with the lowest PI value, which then analyzed by the EASI model. In this work, Monte Carlo technique is used as a stochastic approach to model the uncertainty of Probability of Detection (PD) performance through numerous and different histories of a simulation. Thus, it gives an uncertainty value for PI value that can represent the PPS’s performance confidence level instead of single PI value in the standard EASI model. Also, ASD analysis in this work shows that utilizing other principles, such as covert and rush strategy principle, may result in an intrusion path with the lowest PI value instead of utilizing the CDP concept in some facilities. Furthermore, addition of insider-outsider collusion consideration in ASD analysis of this work, combined with the stochastic- improved EASI model, gives an enriched insight on the effectiveness of the PPS.

Keywords: EASI, Physical Protection System, Adversary Strategy, Insider, Monte Carlo

Topic: Advanced Technology and Other Issues

53 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-86] CRITICALITY INVESTIGATIONS ON EXPERIMENTAL POWER REACTOR WITH THORIUM-BASED NUCLEAR FUEL

Zuhair#, R.Andika Putra Dwijayanto#, Suwoto#, Zaki Su’ud*

#Center for Nuclear Reactor Technology and Safety – BATAN Puspiptek Area, Office Building No. 80, Serpong, Tangerang Selatan 15310, Indonesia *Nuclear Physics and Biophysics Research Group, Department of Physics Bandung Institute of Technology (ITB), Jl. Ganesha No. 10, Bandung 40132, Indonesia

Abstract

Thorium-based nuclear fuel has become an interesting subject for a variety of research with a wide range of applications. Research focusing on thorium-based fuel is aimed to overcome the scarcity and limitation of natural uranium resources as an alternative nuclear fuel in thermal reactor. As thorium has no naturally occurring fissile isotope, it requires other fissile isotope in order to be converted into fissile 233U to produce energy. The isotopes 235U and 239Pu are two of the few alternatives available as the fissile nuclei for thorium-fueled reactor. The purpose of this paper is to investigate the criticality of experimental power reactor (reaktor daya eksperimental, RDE) using two options of thorium-based fuel, namely UO2-ThO2 and PuO2-ThO2. A series of criticality calculations with various uranium contents in UO2-ThO2 fuel and various plutonium contents in PuO2-ThO2 fuel were conducted using the Monte Carlo transport code MCNP6 and continuous energy nuclear data library ENDF/B-VII. CINDER90 depletion module integrated in MCNP6 was utilized in estimating the fuel burnup. The calculation result shows that 35% plutonium content in PuO2-ThO2 fuel has comparable criticality with 80% uranium content in UO2-ThO2 fuel. The former is more effective as nuclear fuel due to larger 233U generation and better neutron economy, thus ensuring longer reactor operational time.

Keywords: criticality, experimental power reactor, thorium-based nuclear fuel, MCNP6, ENDF/B-VII

Topic: Advanced Technology and Other Issues

54 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-87] NEUTRON BEAM CHARACTERIZATION ON THE BEAM TUBES OF G.A. SIWABESSY REACTOR USING MONTE CARLO METHOD

Rasito, Zaki Su’ud, and Sidik Permana

Nuclear Physics Laboratory, Institut Teknologi Bandung, Jl. Ganeca 10, Bandung, Indonesia

Abstract

The neutron beam characterization has been carried out on six tubes of the RSG-GAS reactor for BNCT applications in a simulation using the Monte Carlo method with the MCNP computer codes. The simulation was carried out by modeling the geometry and material of the RSG-GAS reactor with the radiation source model derived from the 235U fission reaction in 40 U3Si2Al fuel bundles with 235U levels of 19.75%. The distribution of neutron and gamma fluences was simulated from the reactor core to the end of the beam tubes S1, S2, S3, S4, S5, S6 with an average tube length of 400 cm and a diameter of 30 cm. The largest neutron fluence is produced by the S5 beam tube of 4.34E+10 cm- 2s-1, also the largest gamma dose of 1362 Sv/j. The smallest neutron fluence produced by the S6 beam tube of 5.9E+9 cm-2s-1, with the resulting gamma dose of 51 Sv/j. Based on the results of the characterization, it was shown that the neutron beam output of each of the RSG-GAS beam tube can be utilized for BNCT applications, but it is necessary to add material to reduce fast neutron fluences and gamma dose.

Keywords: characterization, neutron beam, beam tubes, RSG-GAS , MCNP

Topic: Advanced Technology and Other Issues

55 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-88] EVALUATION OF WIND POWER ENERGY POTENTIAL IN DHAHRAN, SAUDI ARABIA

Gaydaa Al Zohbi (a*), Fatima Ali Wuhayb (a)

(a) Prince Mohammad Bin Fahd university * [email protected]

Abstract

The energy potential of wind in Dhahran city located in the Eastern province of the Kingdom of Saudi Arabia is investigated in the present work. A statistical analysis of recently collected hourly wind data over a period of 7 years between January 2012 and December 2018 at 10 m height by using Weibull distribution function has been presented. The Weibull parameters, shape and scale parameters, to study the wind speed characteristics are used to assess the wind power potential in Dhahran city. In addition, an extrapolation of wind speed and Weibull parameters at different heights (20m, 40m and 60 m) has been carried out. The results are presented as a monthly and annually average for wind speed, shape and scale parameters, probability density function, cumulative distribution function and wind power density. The annual mean wind speed over Dhahran varies from 4.21 to 4.76 m/s at 10 m above ground level. It has been found that the windiest month is June while the calmest is October. The predominant average hourly wind direction in Dhahran is from the north throughout the year. The results indicated that, at 10 m height, the Dhahran city falls under Class 1 (poor) of the international system of wind classification as the mean annual wind speed recorded in the area was 4.35 m/s and the corresponding annual mean power density was estimated to be 68 W/m2. While it falls under class 5 (excellent) at 60 m height as the mean annual wind speed recorded in the area was 6.11 m/s and the corresponding annual mean power density was estimated to be 218 W/m2. Therefore, Dhahran city can be potentially suitable for wind energy applications for a hub height starts from 40 m.

Keywords: Weibull distribution, wind speed, power density, standard deviation method

Topic: Advanced Technology and Other Issues

56 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-90] STUDY ANALYSIS OF TRANSMUTATION SYSTEM USING ACCELERATOR DRIVEN SUBCRITICAL REACTOR Sudarmono (a*), Endiah Puji Hastuti (b*), Anis Rohanda (c*), Andi Sofrany Ekariansyah (d*), Y.Kasezas (e*), Suwoto (f*)

a. Center for Technology and Nuclear Reactor Safety, Gedung 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia, * [email protected] b. Center for Technology and Nuclear Reactor Safety, Gedung 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia, * [email protected] c. Center for Technology and Nuclear Reactor Safety, Gedung 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia, * [email protected] d. Center for Technology and Nuclear Reactor Safety, Gedung 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia, * [email protected] e. Institusi Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran, * [email protected] f. Center for Technology and Nuclear Reactor Safety, Gedung 80 PUSPIPTEK Area, Serpong, South Tangerang, 15310, Indonesia, * [email protected] Abstract

Some countries such as China, Germany, Iran, Poland, Rusia and United State have successfully in designed of Accelerator-Driven Sytem (ADS). Depending on the Subcritical Reactor Assembly, the core size and geometry are very specific and varied for each respective country. The failure of heat removal system of a water-cooled reactor PWR in Three Mile Islands and Fukushima Daiichi BWR made nuclear society consider the use of HTGR for power generation. Therefore, Research and development of new technology for partition and transmutation (P-T) of minor actinide (MA) contained in high-level waste (HLW) has been carried out to design accelerator-based transmutation system or commonly known as Accelerator Driven Subcritical (ADS) device. The objective of this system is to eliminate or minimize high-level radioactive waste that has long half-life and to develop long-term safety assurance in HLW management. Analysis done by using Origen2.1 and MCNP6 codes. As matrix, Th-232 with optimum weight 75% is used.. Comparing with Origen2.1, analysis results using MCNP shows that the use of PWR nuclear spent fuel for ADS device can be done by reprocessing to eliminate U-238 nuclide, which is the source of the formation of plutonium nuclide and minor actinides. Based on the results and discussion of this research, it can be concluded that: Observation on transmutation of transuranic nuclides using ORIGEN2.1 Code on ADS device has been conducted successfully with LMFBR library. 15%-Th232 fuel has been justified as the most optimum ADS fuel based on K-inf reached at EOC on various fuel types. Mass reduction occurs at EOC for Th-232, U-234, U-235, U-236, Np-237, Pu-238, Pu-242, Am-241, and Cm 244. Radionuclides that experience mass changes during cooling time are Pa-233, Pu-238, Pu-240, Am-241, and Cm-244.

Keywords: Transmutation, ADS, Minor actinide, ORIGEN2.1, SRAC2006, MCNP6 Topic: Advanced Technology and Other Issues 57 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-36] NEUTRONIC ANALYSIS OF HCLL-TYPE BLANKETS FOR ITER

Indah Rosidah, Dwi Irwanto, Abdul Waris, Zaki Suud

Institut Teknologi Bandung

Abstract

The viability of blanket fusion reactor design based on the Helium Cooled Lithium Lead (HCLL) blanket is determining using MCNPX in ITER project. Some of the calculation is to achieve TBR ensuring tritium self-sufficiency in effective breeding zone length and to assess the capacity of the blanket shielding under high radiation loading. The effective breeding zone length was depended upon material components of the blanket. For obtaining the tritium self-sufficiency, it was required 80% tungsten enrichment as First Wall (FW) shield in 75 cm blanket region length, and assuming 60% 6Li enrichment in blanket region length of 55 cm. Meanwhile, for evaluating shielding performance, one of the assessment is material using to reduce the radiation loading to the magnetic coil, WC component is good candidate as a vacuum vessel shield.

Keywords: fusion, HCLL, MCNPX, TBR

Topic: Fusion Energy Systems

58 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-37] COMPARATIVE OF BLANKET REACTOR DESIGN IN ASSURING OF TRITIUM SELF-SUFFICIENCY CONDITION

Indah Rosidah, Dwi Irwanto, Abdul Waris, Zaki Suud

Institut Teknologi Bandung

Abstract

There have been performed calculation about blanket module, using either solid, liquid, or molten salt material. The aim of the comparison is not only to find the effective configuration for assuring the tritium self-sufficiency condition, but also to predict the material damage rate and the Helium production rate. In the TBR, the liquid/molten salt breeding material resulted a high TBR (>1.15), whereas Iron material as First Wall (FW) component produced a higher Helium concentration, compared to Cr, Mo, W material.

Keywords: FW, helium, molten salt, TBR

Topic: Fusion Energy Systems

59 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-64] WHAT INERTIAL FUSION R&D CAN DO FOR YOU IN SCIENCE AND TECHNOLOGY

J. Manuel Perlado

Instituto Fusion Nuclear "Guillermo Velarde", Escuela Técnica Superior de Ingenieros Industriales, Universidad Politecnica de Madrid

Abstract

R&D in Inertial Fusion Energy (IFE) follows continuously non-stop in different laboratories along the World, certainly with different acceleration-deceleration many times depending of results of large facilities. The consequence of results from National Ignition Facility (NIF) at Lawrence Livermore National Laboratory drove to what can be interpreted as a slow-down of R&D. That is actually not the reality, because the fiability of the process of development has been increased pushed by a general self-criticism and analysis that make the development more cautious and promising. The European- American connection in the use of OMEGA laser University of Rochester, launching of LMJ-PETAL en Bordeaux, and use of Laser Megajoule (LMJ), the new proposal of facilities closet o Mega-joule in Russia and China, the follow-up of research on Fast ignition in Institute of Laser Engineering (ILE) in Osaka, and the new research on concepts such as “shock Ignition” and a clear renew of high-gain direct drive concept is an evidence of that. In addition, Europe maintains a coordinate research in IFE through the programs of Enable Research. The knowledge in physics for the development of IFE technology, enhanced previously through projects such as European ESFRI HiPER, American LIFE or Japanese KOYO and LIFT, follows with developments in materials for blanket, first wall and optics, Systems and Concepts. Inertial Fusion apart from the Energy benefit for society has more potential spin-off in science and technology. This paper will try to identify some of the beneficial products from this R&D. The codes and development in Radiation-Hydrodynamics has been usefully used to understand several astrophysics questions such as collapsing of supernova remnants, together with the development of extreme ultraviolet sources and X-rays lasers for low dose diagnosis. New technology has been updated for high repetition lasers and use of present ultra-intense lasers. Materials, such as ceramics and nanostructure-based, are now being able for very extreme conditions in the space, thanks to new knowledge in fusion for materials under high radiation, including coatings of engineering systems anti-corrosion. New methods in nanoplasmonic-nanoparticles production appear when solving shielding and filter in optical systems.

Keywords: Energy from Inertial Fusion; Science and Applications Topic: Fusion Energy Systems

60 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-71] STRUCTURAL MATERIAL SELECTION AND ENERGY MULTIPLICATION IN FUSION REACTORS WITH TRANSURANIUM NUCLEAR WASTE

Sümer Şahin (a*), Hacı Mehmet Şahin (b), Hüseyin Şahiner (c), Güven Tunç (d)

(a) Bahçeşehir University,*[email protected] (b) University of Karabük (c) University of Sinop (d) Erciyes University Abstract

Different types of candidate structural materials have been developed and characterized for fusion energy reactors. Among them, steels (austenitic stainless steels and ferritic/martensitic steels), vanadium alloys, refractory metals and alloys (niobium alloys, tantalum alloys, chromium and chromium alloys, molybdenum alloys, tungsten and tungsten alloys), and composites (SiCf/SiC and CFC composites) have primary importance. Steels have unique advantages with respect to extensive technological data base and significantly lower cost compared to other candidates. Furthermore, ferritic steels and modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. However, steels cannot withstand high neutron wall loads to build a competitive fusion reactor. Some refractory metals and alloys (niobium alloys, tantalum alloys, molybdenum alloys, tungsten and tungsten alloys) can withstand high neutron wall loads. But, in addition to their very limited technological data base, they have serious disadvantages due to the high residual radioactivity and prohibitively high production costs. SiCf/SiC composite as structural material in a fusion reactor is attractive based on its low induced radioactivity, low afterheat, high temperature properties and excellent corrosion resistance. However, the improvement of both thermal conductivity and stability of thermo-mechanical properties after irradiation remain the main issue of SiCf/SiC research and development. Also, they are limited with low neutron wall loads despite high temperature resistance up to 1000oC. Conventional thermal reactors, such as LWRs and CANDU reactors are producing substantial quantities of transuranium elements, which represent serious nuisance and permanent hazard potential. On the other hand, they become fissionable material under high energetic fusion neutron irradiation and multiply the fusion energy.In the present work, a Monte Carlo radiation damage analysis has been performed for steels (SS304, SS316, ODS), molybdenum, vanadium and tungsten under consideration of main technical parameters for fusion reactors, such as tritium breeding ratio (TBR), fusion energy multiplication factor (M), displacement per atom (DPA) and gas production (He, H). Numerical calculations have been carried out in spherical geometry with MCNP6 code package using continuous energy cross sections. The 30-neutron-group CLAW-B library was adopted for DPA calculations, while 63-energy groups were used in CINDER calculations for nuclear transmutation of transuranium elements. Structural material selection for the first wall respect to radiation damage limits and optimal reactor performance for reactor blanket have been concluded. Fusion energy multiplication and increase of tritium breeding with nuclear waste transuranium elements are evaluated. 61 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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Keywords: Fusion Reactors; Energy Multiplication; Transuranium Elements; Structural Materials; Material Damage; Tritium Breeding

Topic: Fusion Energy Systems

62 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-89] SAFETY ASSESSMENT OF WENDELSTEIN 7-X EXPERIMENTAL NUCLEAR FUSION FACILITY IN THE CASE OF COOLANT INGRESS INTO VACUUM VESSEL

E. Ušpuras, A. Kaliatka, T. Kaliatka

Lithuanian Energy Institute, Kaunas, Lithuania

Abstract

Wendelstein 7-X (W7-X) is a fusion experiment currently being built in Greifswald, Germany, which shall demonstrate the capability of the stellarator concept on the way to a fusion power plant. As all energy generating devices, this experimental facility must be safe for humans and the surrounding environment. The safety assessment is based on the identification and analysis of the nuclear hazards. The bigger part of radioactive materials is concentrated in the vacuum vessel. Vacuum vessel is the first barrier to prevent the release of radioactive materials into environment. On the other hand, the vessel is designed for the vacuum conditions and pressure slightly above atmospheric could cause the vacuum vessel damage. Any rupture of in-vessel components cooling system pipe (loss of coolant in the facility), leading to ingress of water in the vessel, then may lead to sharp pressure increase and possible damage of vacuum vessel and release of radioactive materials to the torus hall. Lithuanian energy institute (LEI) in the frames of European Fusion Development Agreement (EFDA) program during 2007 -2013 cooperated with Max Planck Institute for Plasma Physics (IPP, Germany) by performing safety analysis of fusion device W7-X. The investigations related to the processes in this experimental stellarator facility did not stopped even later, when LEI in the frame of consortium is participating in EUROfusion H2020 project. In this paper the detailed safety assessment of Wendelstein 7-X experimental nuclear fusion facility in the case of small and large LOCAs is presented. The analysis of response of the cooling system, vacuum vessel and pressure increase protection system during the accident is performed using RELAP5 system computer code. The results of analyses endorsed the suitability of design of experimental facility.

Keywords: Experimental stellarator facility, Vacuum vessel, Thermal-hydraulic parameters

Topic: Fusion Energy Systems

63 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-4] NOISE EXPERIMENTS IN BRAHMMA SUBCRITICAL SYSTEM USING ISOTOPIC POISSON SOURCE AND ACCELERATOR BASED NEUTRON SOURCE

Nirmal Kumar Ray(a,b*), Rajeev Kumar(c), TarunPatel(a), P.S. Sarkar(a,b), L. M. Pant(b,d)

a) Technical Physics Division, Bhabha Atomic Research Centre, Mumbai 400085, India, *[email protected] b) Homi Bhabha National Institute, Mumbai 400094, India c) Reactor Physics Design Division,Bhabha Atomic Research Centre, Mumbai 400085, India d)Nuclear Physics Division,Bhabha Atomic Research Centre, Mumbai 400085, India

Abstract

The energy production and transmutation rate in Accelerator Driven Subcritical (ADS) system depends on the subcriticality of the core. In this perspective, it is essential to develop suitable and robust reactivity measurement methods. Noise methods (Williams, 1974) based on neutron correlation are suitable for reactivity measurement in case of continuous operation without any perturbation. In this context, reactivity measurement using noise methods have been carried out using isotopic Poisson neutron source (Am- Be) and accelerator based non-Poisson neutron source (D-T neutron generator) in zero power subcritical system BRAHMMA, installed in BARC, India for physics studies of ADS (Sinha et al., 2015). In deep subcritical system, reactivity measurement is challenging due to the presence of higher order modes which can be eliminated by placing detectors at locations based on modal analysis (Rana et al., 2013). During noise experiments using mode cancellation method, an Am-Be neutron source (1Ci) has been placed at the centre of the subcritical system (Kumar et al., 2017). However, the proposed external neutron source in ADS is based on particle accelerator is different from isotopic neutron source and the inherent fluctuation in beam current and accelerating voltage makes the former source non-Poisson. Also, periodically chopped beam cannot be considered as Poisson character. In this perspective, Degweker and Rana (2007) had postulated a noise theory based on exponentially correlated Gaussian source characteristics for ADS. In noise experiments based on this postulate, a D-T neutron generator has been used in pulsed mode and source characterisation has been carried out to determine the distribution function, source correlation factor and D+ beam pulse shape. It has been observed that the source is Gaussian in nature, the source correlation factor (50.44ms-1) is very large compared to the prompt neutron decay constant (2.31ms-1) and the pulsed D+ beam is rectangular in shape. Furthermore, the time stamped data have been analysed using various noise methods, namely variance to mean ratio (V/M), Rossi-alpha and Auto Power Spectral Density (APSD) method (Ray et al., 2019). The measured prompt neutron

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decay constant using Poisson source and correlated Gaussian source are in good agreement with theoretical value and amongst them.

Reference 1. Degweker, S. B., and Rana, Y. S. (2007). Reactor noise in accelerator driven systems-II. Annals of Nuclear Energy, 34(6), 463-482.

2. Kumar, R., Degweker, S. B., Roy, T., Singh, K. P., Samanta, S., Bajpai, S., ...and Sinha, A. (2017). Measurement of reactivity in a source driven deep sub-critical system using neutron noise methods. Annals of Nuclear Energy, 103, 315-325.

3. Rana, Y. S., Singh, A., and Degweker, S. B. (2013). Diffusion Theory-Based Analog Monte Carlo for Simulating Noise Experiments in Subcritical Systems. Nuclear Science and Engineering, 174(3), 245-263.

4. Ray, N. K., Pant, L. M., Patel, T., Kumar, R., Roy, T., Sarkar, P. S., ... & Gadkari, S. C. (2019). Experimental validation of noise theory developed by considering the source as an exponentially correlated process in DT neutron generator driven subcritical system. Progress in Nuclear Energy, 117, 103087.

5. Sinha, A., Roy, T., Kashyap, Y., Ray, N., Shukla, M., Patel, T., ...and Adhikari, P. S. (2015). BRAHMMA: a compact experimental accelerator driven subcritical facility using DT/DD neutron source. Annals of Nuclear Energy, 75, 590-594.

6. Williams, M. M. R. (1974). Random processes in nuclear reactors. Pergamon Press, New York.

Keywords: Neutron Noise, Source Correlation, BRAHMMA

Topic: Hybrid Nuclear Energy Systems

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[ABS-15] NEUTRON SOURCES FOR FUSION-FISSION HYBRID REACTORS BASED ON MAGNETIC PLASMA CONFINEMENT : ANALYSIS OF DESIGN PARAMETERS AND TECHNOLOGY READINESS LEVELS

Francesco Paolo Orsitto

CREATE Consortium , Università degli Studi di Napoli Federico II, Napoli(Italy) and ENEA Department Fusion and Nuclear Safety, Frascati(Italy)

Abstract

The aim of the paper is to review the presently available models of neutron sources of fusion-fission hybrid (FFH) reactors based on tokamaks and compare them with other schemes as stellarators and mirrors. Starting from the evaluation of the status of knowledge on Q~1 tokamak neutron sources, the criteria for determining the parameters for a tokamak MCF (Magnetic Confinement Fusion) neutron source are summarized. Using a new scaling law specifically derived for tokamak fusion reactors, plasma parameters for neutron sources useful for FFH based on tokamak are derived. Then a short review of the existing tokamak models already studied for FFH is carried out, comparing the performance predicted by those models with that evaluated using the new scaling law for fusion reactors. The concrete implementation and the radial build of these devices is briefly discussed . The technical readiness level (TRL) of a Q~1-3 MCF tokamak neutron sources is evaluated and R&D lines of research are identified and discussed based on this evaluation .Finally a comparison is made of TRL related to tokamaks neutron sources with other MCF devices like stellarator , mirrors .

Keywords: fusion-fission hybrid , neutron sources, tokamak

Topic: Hybrid Nuclear Energy Systems

66 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-20] COMPARISION OF NEUTRON FLUX DISTRIBUTION OF 233UO2, 235UO2, (TH-233U)O2 AND (TH-235U)O2 FUEL IN THE ACCELERATOR DRIVEN SUBCRITICAL REACTOR

Nguyen Mong Giao (a), Tran Minh Tien (b*)

(a) Center for Nuclear Techniques, Ho Chi Minh City, Viet Nam b) Thu Dau Mot University, Binh Duong province, Viet Nam *[email protected]

Abstract

The idea of Accelerator Driven Subcritical Reactor (ADSR) was mentioned by K. Furukawa et al. [1], C. D. Bowman et al. [2] and C. Rubbia et al. [3]. There have been many studies the structure of the ADSR. M. Hassanzadeh et al. have been simulated TRIGA reactor by MCNPX program [4]. The TRIGA is an ADSR; it contains 102 uranium fuel rods enriched 20% [4]. C.Rubbia et al. have been calculated neutron flux in TRIGA [5].ADSR is very interested and researched because it has many advantages as compared to traditional nuclear reactors; such as: higher safety, the possibility of using various fuels, incinerating radioactive waste and producing energy. To assess the applicability of this technology, the idea of using liquid lead as target and a coolant, the proton beam will interact directly on the liquid lead has been proposed [6,7]. There will be great advantages in this way: there needn’t be the spallation target, instead, we directly use lead not only as the coolant but also the spallation target which interacts with proton beam from the accelerator. So, we will not change the target, manufacture the target and the reactor will not be shut down during the process of operation. The entire lead which is located on the diameter of the reactor will become the spallation target, the length of the target increases and thus the number of neutrons produced also increases.This paper presents results of calculating the neutron flux distribution in an accelerator driven subcritical reactor (ADSR) using UO2, (Th-233U)O2 and (Th-235U)O2 fuel. An ADSR is simulated consists of 90 fuel rods, and 10 graphite reflex rods. All objects are placed in liquid lead. In this paper, thorium is replaced by mixture of UO2, (Th-233U)O2 and (Th-235U)O2; MCNP5 program [8] has been used to simulate and calculate radial distribution of the neutron flux, axial distribution and energy distribution from (p,n) reaction. The calculation results show that the axial distribution of the thermal and fast neutron flux decrease from inside to outside but speed reduction is different. The thermal neuton flux decreases gradually from 0 to 2.5cm; decreases rapidly from 2,5cm to 5cm. The thermal neutron and fast neuron comparision: thermal neutron flux is larger fast neutron flux from 0 to 4cm radius; fast neutron flux larger fast neutron flux from 5cm radius. References [1]. K.Furukawa et al., The combined system of accelerator molten salt breeder (AMSB)

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and molten salt converter reactor (MSCR), Japan–US Seminaron Thorium Fuel Reactors, Naora, Japan, October 1982.

[2]. C.D.Bowman et al. (1992), Nuclear Instruments and Methodsin Physics Research A320, 336–367.

[3]. C.Rubbia, et al., CERN/AT/95-45(ET)/(1995).

[4]. M. Hassanzadeh , S.A.H. Feghhi , Sensitivity analysis of core neutronic parameters in accelerator driven subcritical reactors, Annals of Nuclear Energy, Elsevier, 2014.

[5]. C.Rubbia, et al., Preliminary neutronic analyses of the TRIGA-ADS demonstration facility, Korea, 2002.

[6]. N. M. Giao, V. T.D.Hang , T. M. Tien, Ability to Make Accelerator-Driven Sub-Critical Reactor System (ADS) Without A Separate Spallation Target for (p,n) Reaction, International Journal of Modern Physics and Application, 2015.

[7]. Tran Minh Tien, Distribution of Neutron Flux from (p, n) Reaction on the Liquid Lead Target for Accelerator Driven Subcritical Reactor (ADSR) , J. Phys.: Conf. Ser. 1172 012066, 2019

[8]. X-5 Monte Carlo Team, MCNP – A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos Nation Laboratory, 2003

Keywords: ADSR, subcritical, neutron flux, thorium fuel

Topic: Hybrid Nuclear Energy Systems

68 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-25] BURNING OF MINOR ACTINIDES IN STELLARATOR-MIRROR FUSION- FISSION HYBRID

S.V. Chernitskiy (a), V.E. Moiseenko (a), O. Agren (b)

a) National Science Center Kharkiv Institute of Physics and Technology. Kharkiv, Ukraine b) Uppsala University, Uppsala, Sweden

Abstract

The stellarator-mirror fusion-fission was proposed for spent nuclear fuel transmutation and energy production [1]. The calculations were done for the MOX fuel containing plutonium and minor actinides from the spent nuclear fuel. In the present report, the possibility to use the same fuel, but without plutonium is analysed. The MCNPX numerical code has been used to model the neutron transport in a sub-critical fast fission reactor driven by a fusion neutron source. Heat load on the first wall, the distribution of the neutron fields in the reactor, the neutron spectrum and the distribution of energy release in the blanket are calculated. The possibility of tritium breeding inside the installation in quantities that meet the needs of the fusion neutron source is analyzed. The calculations indicate that usage of the minor actinide fuel is possible in such a hybrid reactor.

1. V. E. Moiseenko, et al., Plasma Phys. Control. Fusion 56 (2014) 094008 (11pp).

Keywords: Fusion, Fission, Stellarator, Fast reactor, Hybrid

Topic: Hybrid Nuclear Energy Systems

69 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-57] RESEARCH AND DEVELOPMENT ON ADS FOR MINOR ACTINIDES TRANSMUTATION AND THORIUM UTILIZATION

Marcia S. Santos(1), Renato V. A. Marques(1), Felipe Martins(1), Graiciany de P. Barros(2), Carlos E. Velasquez(1), Antonella L. Costa(1), Maria Auxiliadora F. Veloso(1) and Claubia Pereira(1)

1Departamento de Engenharia Nuclear - Universidade Federal de Minas Gerais (1)Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Presidente Antônio Carlos, 6627 campus UFMG 31.270-901, Belo Horizonte – MG, Brazil

(2) Centro de Desenvolvimento da Tecnologia Nuclear Av. Presidente Antônio Carlos, 6627 campus UFMG 31.270-901, Belo Horizonte – MG, Brazil

Abstract

The nuclear engineering department (DEN) at UFMG has been working with reprocessing techniques since mid-90’s. The recycling and reprocessing technique were investigated at DEN/UFMG with the purpose to recover fission products and the actinides produced in thermal reactors with long half-life. On the one hand, transmutation induced by fission could reduce the minor actinides half-life, to achieve higher fission probability it is needed a hardened spectrum. An Accelerator Driven System produces neutrons with high energy through spallation reactions of the protons with a cylindrical lead target. On the other hand, to avoid transmutation induced by uranium radiative capture increasing the number of minor actinides, it was proposed to use thorium, which is a fertile material producing 233U. Therefore, the DEN/UFMG has been focus on study the GANEX reprocessed technique spiked with thorium and how the variation in fissile material are important in the processes of fission and production of fertile material. Other studies using different reprocessing techniques has been implemented showing the advantages of each reprocessing, as well as, the neutron behavior. In addition, the system has also been used to test a fusion source instead of the spallation source to see the difference in transmutation by change the neutron source. Finally, using the conventional spallation source, different coolants such as NaK, Na, LiPb, LBE and Pb have been tested to study the neutronic parameters, the neutron spectrum and the transmutation influence over the nuclear fuel during 10 years of burnup. Therefore, this work presents a summary of the work developed at DEN/UFMG with ADS.

Keywords: Thorium; ADS;transmutation; hybrid system Topic: Hybrid Nuclear Energy Systems

70 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-58] PROPOSAL OF A FUSION-FISSION SYSTEM BASED ON THE ITER PROJECT

Renato V. A. Marques, Marcia S. Santos, Felipe Martins, Carlos E. Velasquez and Claubia Pereira

Departamento de Engenharia Nuclear - Universidade Federal de Minas Gerais Av. Presidente Antônio Carlos, 6627 campus UFMG 31.270-901, Belo Horizonte – MG, Brazil

Abstract

The nuclear engineering department (DEN) at UFMG has been studied hybrid system such as the fusion-fission system (FFS). The study of the FFS begins modeling a Tokamak evaluating the neutronic characteristics of the first wall materials such as beryllium and tungsten alloys and the combinations of them as a plasma facing components (PFC). This study concluded that the most suitable combination is to place the beryllium alloy in the external PFC to have a hardness spectrum. Then, the analysis was focus on the neutron flux along the system to localize the best position to place a transmutation layer inside the Tokamak, indicating that it was between the heat sink and the block shield. After that, several transmutation layer thicknesses were tested, between 15 to 40 cm, using LiPb as a coolant. Two nuclear fuels were studied, both of them reprocessed by UREX+ technique with 20% of fissile material, but one of them spiked with thorium and the other with depleted uranium. The results show that the optimal transmutation layer thickness is 20 cm because enhances the transmutation of the transuranic elements. Nonetheless, to enhance transmutation from the modeled system, different liquid metal coolants have tested reducing the amount of fissile material to 11.5% due to the neutronic parameters change for each coolant. The best choice considering transmutation by fission reactions was the LEB coolant that in detriment, it has lower tritium production. Therefore, the following studies focus on the insertion of a tritium breeder layer before the transmutation layer to enhance the tritium production and then analyzed the implications over transuranic transmutation. In order to continue the research, different reprocessing techniques have been studied trying to applied all the improvement already development with the purpose to enhance transmutation on the system.

Keywords: fusion-fission system; transmutation, nuclear hybrid system

Topic: Hybrid Nuclear Energy Systems

71 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-61] CASE STUDY OF A NUCLEAR-SOLAR PV HYBRID BASE LOAD PLANT

A. K. Mathur, V. Jain, S. A. Khan, V. Jagannathan and Suneet Singh

Indian Institute of Technology, Bombay Bhabha Atomic Research Centre Atomic Energy Regulatory Board Insolare Energy Pvt Ltd.

Abstract

The twin problems with wind and solar power plants is that (a) they cannot vary their output in response to changes in demand and (b) their generation is intermittent. To overcome both these limitations, it is proposed to couple them with either a fossil or a nuclear power plant and treat the coupled plant as a pseudo-base load plant with a single connection to the grid. The objective is to fully utilise the installed capacity of the renewable energy plant while smoothing the variability of their energy output. The power level of the coupled fossil/nuclear power plant can be varied so that the sum of the outputs of the coupled plant is almost flat throughout the day. This helps the grid to draw power from clean renewable sources without any adverse effect on its stability. In this paper the feasibility of balancing the variable power output of a solar power plant by suitable power maneuvers from a nuclear power plant is examined. Although the energy from a solar PV plant is variable, it can be predicted with a high degree of accuracy. We can take advantage of this predictive capability to pre-program the power ramps of the nuclear plant to offset the hourly power generation from the solar power plant. The power level of the nuclear power plant is varied by changing in soluble boron concentration since this does not disturb the axial power shape. Numerical studies were carried out to ascertain the capacity of the PV solar power plant that can be used as a companion plant for a 1000 MW reactor. The analysis showed that a 500 MWe solar PV plant can be coupled with a single 1000 MWe VVER without any modification to the existing plant design. The pseudo-base load plant addresses the problem of intermittency of renewable energy and increases its share in the energy mix without affecting the stability of the grid.

Keywords: Solar nuclear hybrid, base load, renewable energy

Topic: Hybrid Nuclear Energy Systems

72 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-65] FUSION-FISSION HYBRID SYSTEM EXPERIMENT BASED ON A TRIGA REACTOR IN A SUB-CRITICAL CONFIGURATION

Fabio Panza, Mario Carta, Marco Ciotti, Nadia Cherubini, Alessandro Dodaro, Valentina Fabrizio, Luca Falconi, Francesco Filippi, Luigi Lepore, Francesco Orsitto, Mikhail Osipenko, Mario Palomba, Giovanni Ricco, Marco Ripani, Massimo Salvatores

- ENEA Casaccia S. Maria di Galeria - Roma (Italy) - CREATE Consortium , University di Napoli Federico II, Napoli - Italy and ENEA Fusion and Nuclear Safety , Frascati - Italy - Istituto Nazionale di Fisica Nucleare - Sezione di Genova, Genova, Italy - Centro Fermi - Museo Storico della Fisica e Centro Studi e Ricerche “Enrico Fermi’, Rome, Italy e Senior Scientific Advisor

Abstract

A hybrid fusion-fission reactor is a potential option for integrated systems from one side for energy production and from another side for fission-reactor radioactive-waste transmutation. In order to preliminary investigate the neutronics characteristic of such a system we are envisaging the possibility to set the TRIGA-RC1 research reactor (ENEA- Casaccia Research Center) in a low-power sub-critical mode configuration, and fed by a d-t neutron generator as a driving source. Provided a sufficiently high neutron flux can be achieved, innovative diamond neutron detectors may be used to measure neutron energy with resolution about 100 keV. In this paper it will be addressed the issue of the coupling optimization, the possible experimental program, and the identification of a suitable neutron detectors arrangement. These experiments could contribute to preliminary design, uncertainty reduction, and to the validation of related nuclear data and simulation codes.

Keywords: Fusion, fission, hybrid systems, subcritical systems

Topic: Hybrid Nuclear Energy Systems

73 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-77] COUPLING ELECTRON ACCELERATORS TO HEAVY WATER REACTORS: PHOTON TRANSPORT AND PHOTONEUTRON PRODUCTION

Silvia da Costa Frias de Barros1, Gitae Kim2, Yeseul Seo2, Douglas A. Fynan2

1. Sejong University 2. Ulsan National Institute of Science and Technology (UNIST)

Abstract

Small-scale photoneutron sources and neutron beams have been proposed for radioisotope production and boron neutron capture therapy by coupling commercially available electron linear accelerators (eLINACs) to beryllium or deuterium systems. This work investigates the coupling of eLINACs to large heavy water reactors to create intense photoneutron sources from the conversion of bremsstrahlung photons into photoneutrons in the heavy water moderator. The photoneutron sources maybe useful in practical heavy water reactor operational applications including local (zone) reactivity control, flux flattening, Xenon override by subcritical multiplication, dampening Xenon oscillations, source range detection, and low power operation. In large heavy water reactors that are spatially decoupled, local photon production in a bremsstrahlung converter, photon transport and photoneutron production need to be well understood before the engineering applications of the hybrid system can be explored. This paper presents analytical and numerical simulation results for the important photon radiation transport phenomena that influence photoneutron production in a CANDU6 fuel channel lattice comprised of large volumes of heavy water (γ,n) converter material between widely spaced fuel channels which are a strong photon attenuating material. The important phenomena are energy loss during Compton scattering in relation to the deuterium photoneutron reaction threshold energy, pair production competition at high photon energies, and importance of modeling secondary and tertiary radiation particles when photon energy is greater than 10 MeV. Photoneutron yield data are produced for monoenergetic photons between 2.5 MeV and 20 MeV and representative bremsstrahlung spectra from eLINAC beam powers of 5 MeV, 10 MeV, and 20 MeV. Electron transport and bremsstrahlung production was performed with the MCNP6 code and all photon transport calculations in the CANDU6 model were performed with the MCS code.

Keywords: accelerator-driven system, photoneutrons, heavy water reactor

Topic: Hybrid Nuclear Energy Systems

74 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-14] CLIMATE AND HEALTH BENEFITS FROM THE EXPANSION OF NUCLEAR POWER

Minghuang Wang, Xuewei Fu, Chao Lian, Dehong Chen, Zhibin Chen, Yunqing Bai, Fang Wang, Liqin Hu, Yican Wu*, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031, China

Abstract

In recent history, nuclear power played a significant role in climate change mitigation, and it contributed 41% of all avoided emission by all low carbon energy sources in last 50 years. Nuclear power is also much cleaner than fossil fuels with almost zero emissions of air pollutants. It helps us attain multiple aspects of the goal of sustainable development all over the world. To achieve the 2 °C target in Paris Agreement, fossil fuel energy must be partly substituted by low-carbon energy in the future. Nuclear power is an essential and feasible approach for sustainable future that will reduce carbon emissions and air pollutions and build environment resilience. However, nuclear power’s expansion endures a lot of controversy regarding its health risk from radiation. It is thus necessary to make a comprehensive analysis to quantify the climate and health benefits of nuclear power. Here we show the regional and global impacts of nuclear power on climate and human health. Using the Representative Concentration Pathways (RCPs) without nuclear as a basis for climate predictions and projections, we show that only applying International Atomic Energy Agency (IAEA) nuclear high projection in the RCP2.6-without nuclear scenario could extend the time to reach the 2°C warming beyond 2100. A sensitivity analysis by scanning the nuclear share of electricity generation reached in 2100 is also conducted. We also demonstrate that, based on three nuclear power projections from 2016 to 2050, the projected global use of nuclear power would prevent 1.7-14.0 million deaths that would result from coal power, and the monetized benefits range from US$16.7 to 115.0 trillion. In particular, less developed regions will benefit more from the replacement of coal by nuclear power. Nuclear power would act as a stabilizing force in the global sustainability.

Keywords: Nuclear power, Global warming, Human health, Air pollution, Radiation

Topic: Knowledge, Management, Human Resources and Social Issues

75 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-17] DEVELOPMENT OF HEXAGONAL-Z GEOMETRY CAPABILITY IN RASTK FOR FAST REACTOR ANALYSIS

Tuan Quoc Tran, Alexey Cherezov, Xianan Du, Jinsu Park, Deokjung Lee*

Ulsan National Institute of Science and Technology (UNIST) Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Ulsan, 44919, Republic of Korea. *Corresponding Author: [email protected]

Abstract

Recently, sodium-cooled fast reactor developments have been active with experimental and research reactors [1]. It is considered as one of promising reactor types that can meet the Generation IV International Forum (GIF) goals. The three-dimensional (3D) two- group nodal diffusion code RAST-K v2.0 has been developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) with the initial application goal to Light Water Reactors (LWRs). It was successfully verified and validated that it can provide results in good agreement with measured data of LWRs[2]. In other to use it for Sodium cooled Fast Reactor (SFR) cores, the RAST-K code is under further development for the hexagonal-z geometry (RASTK-HEX) including the extension of two-group rectangular solver to multi-group hexagonal solver and the update of thermal-physical properties of fast reactor core materials in the internal thermal-hydraulic solver. To solve the multi-group neutron diffusion equation in the 3D hexagonal-z geometry, the triangle-based polynomial expansion nodal (TPEN) method is implemented in the RASTK-HEX code [3]. The capability of RASK-HEX to perform steady state analyses of SFR cores is demonstrated in this paper using benchmarks of sodium-cooled fast reactor cores with various fuel types and core sizes (MOX-3600, CAR-3600, MET-1000, MOX-1000) [4]. The paper will describe RASTK-HEX development and present its results of the steady state simulation of benchmarks.

Keywords: sodium-cooled fast reactor, SFR, nodal diffusion code, RAST-K, TPEN

Topic: Knowledge, Management, Human Resources and Social Issues

76 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-19] PHASED AND DYNAMIC FAILURE MECHANISMS CONSIDERED RELIABILITY ANALYSIS OF THE DIESEL GENERATOR SYSTEM AFTER LOOP

Daochuan Ge, Zhen Wang, Shanqi Chen, Chao Chen, Zhibin Chen*, Fang Wang, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China

Abstract

After loss of offsite power (LOOP), loss of power supply from the diesel generator system (DGS) is a very severe accident whose failure behaviors can be divided into two stages according to reparability, and poses great threat to the safety for both old and new nuclear power plants (NPP). Hence, it is necessary to make an exact reliability analysis of the DGS. The traditional analyzing tools to evaluate the reliability of NPP systems are static approaches, such like fault tree, event tree and etc. Such static modeling tools are extensively applied in NPP due to their simplicity and integrated analyzing techniques. However, static modeling techniques are not capable of capturing the dynamic sequence-dependent failure behaviors which typically exist in NPP systems and often overstate the unreliability of systems due to ignoring the dynamic failure behaviors. In this paper, we take two phased dynamic fault trees (DFT) to model dynamic failure behaviors of DGS after LOOP. In the first phase, the system is considered to be non- repairable, and in the second phase, the system is considered to be repairable. An integrated Markov Chain model is proposed to calculate the system’s reliability. Comparative study between DFT and static fault tree (FT) is also carried out. The results indicate that unreliability of DGS calculated by static FT is greatly overstated compared with those derived from the DFT model, and hence it is necessary to take DFT model to analyze NPP safety system with dynamic failure behaviors.

Keywords: phased mission system; dynamic reliability analysis; dynamic fault tree

Topic: Knowledge, Management, Human Resources and Social Issues

77 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-24] TOOLS TO EDUCATE, AND DISPEL MYTHS AND MISCONCEPTIONS

Zahra Hanifah and Rizwan-uddin

University of Illinois

Abstract

Efforts to educate the general public about radiation in general and nuclear energy in particular and thus dispel misconceptions are needed. Researchers at the Virtual Education and Research Laboratory (VERL) at the University of Illinois develop tools— such as 3D interactive models, VR/AR applications and video games—for education and training of students and workers, as well as to educate non-professionals about concepts in radiation protection, shielding, and nuclear reactor technology. Three-dimensional (3D) computer based models and games have been created to educate, train, and introduce these concepts—specifically a virtual, immersive tour of a research reactor, classroom laboratory experiments including virtual experiments to measure half life of a radioactive substance and the attenuation coefficient of shielding materials, and scavenger hunt game to be played in a radiation environment where the goal is to minimize the dose received during the game. These 3D models have been presented to people of all ages during open houses, lab tours and campus visits on platforms including desktop and laptop computers, AR devices such as Oculus-Rift and Oculus-Go, and mobile phones. Some of these models have also been used in courses as well. The specific models described and discussed in this paper each have goals and targets in teaching users in specific ways and have displayed levels of success in doing so. The computer based classroom laboratory models provide a straightforward teaching approach to topics such as measurement of specific heat of metals and attenuation coefficient of shielding material. The user is able to mimic real laboratory experiments including measuring data such as temperatures, flow rates, and radiation count rate, and interacting with realistic (virtual) equipment. The fully immersive tour of UIUC’s (now decommissioned) TRIGA reactor is an interesting and efficient way to introduce people to nuclear facilities. This model also demonstrates a potential use of VR as workplace training tool. 3D model of the TRIGA facility has also been used to develop a scavenger hunt game. Users are to navigate the facility and collect objects while minimizing dosage, thus learning about key concepts in radiation protection in doing so. These methods for outreach and education have proven to be extremely engaging and impactful especially to those in the middle/high school age range. Additionally, these models have also shown the potential to be used as training tools at nuclear facilities.

Keywords: Knowledge Management, Outreach, Training and Education, Virtual and Augmented Reality

Topic: Knowledge, Management, Human Resources and Social Issues 78 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-30] THE DEVELOPMENT OF MINIATURE NUCLEAR REACTOR PROPS AS A LEARNING MEDIA

Casmika Saputra (a), Abdul Waris (b*)

a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia b) Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia

*[email protected]

Abstract

Nuclear reactors are used to produce and control the release of energy from sustainable atomic chain reactions. For a nuclear power plant, the heat energy produced by the nuclear reaction is then used to generate electricity. The main components for controlling the nuclear reactions are fuel, control rods, moderator, and coolant. Each has roles and functions which in teaching generally explained through graphic illustrations as learning media. Almost or even never the nuclear reactors are demonstrated using props. As we know, props could be exciting and effective learning media. Therefore, at this work, we develop miniature nuclear reactor props as learning media. It consists of a miniature nuclear reactor model and a friendly graphical user interface (GUI). So, learners could manage the reaction process in nuclear reactors through the GUI. However, the actual nuclear reaction is not possible to use in this props. Therefore, as a substitute for the heat produced by the nuclear reaction, in the miniature nuclear reactor model, heat is made using electric heating elements (heater). With the props, students are expected to be easy to understand the working mechanism of nuclear reactors in general.

Keywords: Learning Media; Nuclear Reactor; Props

Topic: Knowledge, Management, Human Resources and Social Issues

79 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-45] MULTILATERAL APPROACH TO THE NUCLEAR FUEL CYCLE: AN ALTERNATIVE AND THE MALAYSIAN PERSPECTIVES

Bashillah Baharuddin, Abdul Muin Abdul Rahman and Siti A’iasah Hashim

Malaysian Nuclear Agency Ministry of Energy, Science, Technology, Environment and Climate Change Kompleks Bangi 43000 Kajang Selangor, Malaysia

Abstract

Due to the dual use nature of nuclear technology, the policy debate on nuclear technology especially the nuclear power program, touches on a very sensitive political topic in the context of the ongoing ‘war on terror’ specifically ‘nuclear terrorism’. To prevent states from misusing sensitive technology such as the enrichment and reprocessing of nuclear materials for proliferation purposes, the nuclear supplier states had suggested a new approach called the multilateral nuclear arrangements (MNA). However, the MNA has come under criticism, especially from the developing countries, since it contradicts state rights for peaceful uses of nuclear technology, as stipulated under Article IV of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). This paper explores possibility of Malaysia’s participation in the MNA, contributing to the debate on the most appropriate option for nuclear fuel cycle and provides information for developing Malaysia’s Nuclear Fuel Cycle Policy in the future.

Keywords: MNA, NPT, non-proliferation, nuclear fuel cycle policy

Topic: Knowledge, Management, Human Resources and Social Issues

80 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-78] ENERGY FOR OUR FUTURE: A SYNERGETIC APPROACH BETWEEN SHORT- TERM NUCLEAR FISSION AND LONG-TERM NUCLEAR FUSION

J.-M. Noterdaeme

Ghent University, Ghent, Belgium

Abstract

The use of fossil fuel must be urgently curtailed to avoid drastic consequences for the climate of our planet. To meet the CO2 targets, all new power plants must be CO2 free. This means that every power plant newly installed to meet our world-wide drastic increasing demand in energy, and any replacement of an ageing power plant, must be CO2 free. Renewable energies, such as solar and wind, are partially an option. But renewables are presently not and most likely will also in the future not be sources of reliable, always on-demand available power, unless substantial progress is made in storage capabilities or geographically very distant regions are interconnected. Nuclear fission, in its historical form with one-of-a-kind large power plants that have been developed to take advantage of economies of scale, are now beset by problems related to their large scale. Among those problems are very long building times with often large cost-overruns and the limited number of industrial actors able to build the large scale components needed. In addition, the lack of public acceptance and institutionally shifting conditions for plant operation make the utilities very reluctant to invest in them. Nuclear fusion could be a solution, but it is extremely unlikely that fusion power plants can be deployed in time to provide the urgently needed CO2 free power.

How do we solve this dilemma? To build and install sufficient numbers in a short time, nothing beats a standardised factory-built approach (like for wind power generators, cars, airplanes, …). Small modular fission power plants can be factory built. They have a number of advantages with respect to their large one-of-a-kind counterparts. Think of their manufacturing and licensing in the same way as it is done for airplanes. Being factory built, they can experience better quality assurance, enjoy economies of numbers with reliable cost estimates and on-time delivery, and be pre-licensed. Being small, they can be added in increments as capacity is needed; this results in smaller investment steps. The load can be followed in a more efficient fashion, by turning off some reactors, rather than by operating a large one at fractional load. Maintenance can be rotated among a set of small reactors, thereby taking out of the network only a small fraction of the total power. With a larger surface to volume ratio, small reactors are easier to cool in the case of an accident, which make them safer. How do we tackle the lack of public acceptance? We argue that this is mostly due to the waste problem. The still hanging questions of whether the waste should be reprocessed or not, retrievable or not, compound the issue. Only some countries have found solutions for disposing of the nuclear waste. In most countries, the waste is still temporarily stored at the power plants 81 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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while they are in the process of finding sites for the repositories. If nuclear fission were the long-term solution, storage will have to grow linearly with our energy consumption. Reprocessing may face even stronger public opposition than waste storage. Immediate reprocessing may change the amount that needs to be stored but will not change the continued increasing need for storage. Making the waste retrievable, motivated by the wish to reprocesses it in the future, when better reprocessing methods become available, or motivated by the need to access non-used fuel, if fuel becomes scarce, makes building and maintaining a waste repository for the needed long time scales unbelievable. If one can rely on fusion for the long-term solution of our energy problem, then the boundary conditions for the use of nuclear fission changes as it is only a short- term solution. There is no need for reprocessing, neither to reduce the amount of waste that will be limited, nor because of concerns about fuel availability. There is no requirement to store the waste in a retrievable way. A once-through use of the uranium fuel, without reprocessing and with a permanent and definitive disposal of the waste, will go a long way towards public acceptance. Therefore, we propose to use small modular nuclear reactors with once-through fuel cycle and permanent disposal of the waste to care now of our short-term energy needs, while relying for the long-term on nuclear fusion.

Keywords: energy policy, modular reactors, nuclear fusion

Topic: Knowledge, Management, Human Resources and Social Issues

82 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-94] WHY INDONESIA SHOULD DEVELOP NUCLEAR POWER?

Muhammad Busyairi

Completed Master in Public Administration (Policy) at Flinders University, South Australia in August 2019, now returned to his institution, West Lombok Government, Indonesia [email protected] or [email protected]

Abstract

Energy policy in Indonesia is still not in line with what is being argued for by the global community today: the efforts to cope with climate change. During past decades, the amount of carbon-dioxide (CO2) in the atmosphere has increased significantly due to the more people using coal, oil and natural gases. As a big country Indonesia is with the population of more than 260 million, Indonesia needs a great amount of electricity supply. While most of the energy source is from fossil fuels sources such as coal, oil, and gas, the need for energy increases gradually. However, the reliability of the electricity supply is questionable with many protests due to rolling blackouts. Even, President Joko Widodo admitted that there have been many complaints from the public about the frequency of power failure. One of the most promising alternatives to cope with energy demands and greenhouse gas emission is by developing nuclear power plants. Ironically, although Indonesia has planned to develop nuclear power since 1950s, it has not been really developed till nowadays for some reasons. This research by using a qualitative approach and based on data taken from books, e-books, journal articles, and governments’ documents aims to analyze the need for building nuclear energy in Indonesia. It has four findings and suggestions: (1) nuclear power is the best option for Indonesia because it is safe, can meet energy demands, and economical. (2) nuclear power is the best option to mitigate and adapt global warming as suggested by Giddens and eco-modernists since it is environmentally friendly. (3) while the fear of lay people towards nuclear power development is the true fear because of unknowingness, the educated opponents and political elites are “playing the politics of fear” to bring pessimistic and despair. And (4) the President should not hesitate to make policy to develop nuclear power because it has been widely supported by Indonesian public.

Keywords: nuclear power, climate change, eco-modernist, politics of fear

Topic: Knowledge, Management, Human Resources and Social Issues

83 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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[ABS-12] INTELLIGENT FAULT DIAGNOSIS METHOD AND COMPUTER AIDED REGULATION SYSTEM IN BEAM COMMISSION OF HINEG FACILITY

C. Zhao(a,b), J. Wang(a), Y. Wang(a), Z. Wang(a), Q. Zhang(a), Y. Zhang(a*), FDS Team(a)

a) Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China *yong.zhang2 @fds.org.cn b) University of Science and Technology of China, Hefei, Anhui, 230031, China

Abstract

High Intensity D–T fusion Neutron Generator (HINEG), developed by the Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences, is an essential platform for advanced nuclear energy systems and nuclear technology applications. HINEG has produced the D-T fusion neutrons with the yield of 6.4\times10^{12} n/s. Now, the upgrading of HINEG to reach a higher neutron yield more than 10^{13} n/s is on-going. To achieve this target, the higher stability and reliability of the hardware and software systems are required. A method of digital intelligent fault diagnosis and the computer aided regulation system are applied in the commission process for the ion source and the neutron source. The realization of digital intelligent fault diagnosis in HINEG is based on the full digital hardware and software architecture, and the dual redundancy intelligent power supply debugging mode. The system of computer aided regulation can be used to predict the possible faults, such as high-voltage sparking, vacuum anomaly, beam dropout and beam bias, and so on. It can give operating parameter suggestions for the neutron source in the process of beam commission by combing the real-time data and the historical data. The results of beam commission experiments indicate that the method of digital intelligent fault diagnosis and the computer aided regulation system help to improve the reliability and stability of HINEG facility.

Keywords: HINEG, Intelligent fault diagnosis, Computer aided regulation system

Topic: Nuclear Energy Expanded Applications

84 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-35] STUDY HEAT TRANSFER IN NATURAL CIRCULATION OF LIQUID SODIUM FOR STEADY STATE AND TRANSIENT CONDITIONS

Rindi Wulandari(1); Sidik Permana(2); Suprijadi (3)

(1)(2)Nuclear Physics and Biophysics Research Division, Institut Teknologi Bandung Jl. Ganesha 10, Bandung 40132, Gedung Fisika FMIPA ITB Indonesia

(3) Theoretical High Energy Physics and Instrumentation Division, Institut Teknologi Bandung Jl. Ganesha 10, Bandung 40132, Gedung Fisika FMIPA ITB Indonesia

1) [email protected] (corresponding author) 2) [email protected] 3) [email protected]

Abstract

One of the problems in fullfing energy needs in Indonesia is marked by the low electrification ratio, which is 60%. Many researchs dan various studies of alternative energy has been conducting to solve these problems. One of them is nuclear energy. The development of nuclear power plant (NPP) is very rapid. Nowdays, many studies of 4th Generation nuclear reactor which focus on improving safety is conducted. The characteristic of some IV generation nuclear reactors is the use of molten salt as a coolant. The purpose of this study is to determine the heat transfer of molten salt in the natural circulation system for steady state analysis and transient characteristic with COMSOL Multiphysics method. The selected module is the Non-Isothermal FLow (NITF) module. This module is a combination of three basic equations, namely the continuity equation, the Navier-Stokes equation, and the dynamic equation of heat transfer in fluid. The simulation model measures 1.5 x 2 (m) with sodium (Na) as a fluid. The simulation demonstrates 5 conditions: 1) Steady state; 2) Transient I; 3) Transient II; 4) Loss of Heat Sink; 5) Heat Trip. The obtained results are the form of temperature distribution, pressure, and velocity of fluid flow from each state.

Keywords: heat transfer, natural circulation, COMSOL Multiphysics method

Topic: Nuclear Energy Expanded Applications

85 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-46] NEUTRONIC DESIGN OF 100 MWE MSR WITH TH-PU-MA FUEL

Cici Wulandari(a) , Abdul Waris(b*), Robi Dany Riupassa(a), Sidik Permana(b) , and Yazid Bindar(c)

a)Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA b)Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung,INDONESIA c)Department of Chemical Engineering, Faculty of Industrial Engineering, Institut Teknologi Bandung,INDONESIA *Email: [email protected]

Abstract

Molten Salt Reactor (MSR) has been consider as a nuclear reactor system that is quite promising for Indonesia due to its advantages in the inherent safety, non-proliferation, and sustainable energy systems. We have studied a 100 MWe MSR with a homogeneous reactor core consisting of molten salt as fuel and graphite as a moderator. This reactor will be operated for 5 years without refueling. The molten salt used is FLiBe (LiF-BeF2) in eutectic state with Th-U and Th-Pu-MA. The performance of reactor will be studied by evaluating some of the neutronic parameters, such as the effective multiplication factor, conversion ratio, neutron flux and atomic density of each nuclide. Neutronic design were performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 4.0 nuclear data library.

Keywords: Minor Actinide, Plutonium, Uranium, Thorium, MSR, SRAC 2006, JENDL 4.0

Topic: Nuclear Energy Expanded Applications

86 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-49] COMPARATIVE STUDY OF DIFFERENT MONTE CARLO METHODS FOR MULTI-GROUP NEUTRON DIFFUSION CONSTANTS GENERATION

Fatemeh Mohammadhasani, Mohammadreza Nematollahi*

Shiraz university

Abstract

Monte Carlo (MC) method is one of the most efficient methods to generate the multi- group diffusion constants (MGDCs) for the advantage of complex geometry description and use of continuous energy cross sections data library. The main goal of this paper is the implementation and comparison of different methods to generate MGDCs such as scattering cross section and diffusion coefficient using MC method (specifically in this case the MCNP code). To this, first the different methods are implemented based on the MCNP MC code as a powerful calculation tool due to its high accuracy in neutronic computations. Then three-groups parameters resulted from the implementation of different methods in the MCNP code are compared for a pin-cell benchmark problem. To evaluate the results, WIMSD5 deterministic code is used as reference. Also, in this study two group constants obtained from the simulation results of the performed methods in the MCNP MC code for two practical problems, the fuel elements of Tehran research reactor (TRR), are compared. The applied methods in the MCNP MC code can be used to generate group constants for use in core calculations codes such as PARCS code for the steady state calculations and reactivity transient simulations.

Keywords: Multi-group diffusion constants generation, Monte Carlo method, MCNP code

Topic: Nuclear Energy Expanded Applications

87 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-52] ANALYSIS OF NEUTRONIC AND NON-PROLIFERATION ASPECTS WHIT FUEL VARIATIONS ON MOLTEN SALT REACTOR (MSR) BASED ON THORIUM FUEL

Ade Maulana Fadillah(a*), Sidik Permana (b)

a) Master Program of Physics Department, Faculty of Mathematics and natural science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung 40132, Indonesia *[email protected] b) Nuclear Science and Engineering Dept., Physics Dept., Bandung Institute of Technology

Abstract

Molten Salt Reactor (MSR) is one of the design concepts of the fourth generation reactors that have been developed. Liquid fuel or molten salt fuel is used for MSR based on uranium and thorium fuels. The purpose of this research is to analyze neutronic and non-proliferation aspects of the MSR based on reactor type of FUJI U3 as MSR thermal reactor type and molten salt fast reactor (MSFR) as fast reactor type. An established computer code of SRAC 2006 and adopted nuclear data library of JENDL 4.0 are used for this evaluation. Some evaluations have been done for different temperature levels and composition of fuel for MSR FUJI-U3 and MSFR until the condition for reactor criticallity and the fuel composition is analyzed for non-proliferation aspect evaluation. From the obtained results, for both types of MSR (MSR FUJI-U3 and MSFR), increasing fuel temperature can reduce the criticality condition and increase the value of the conversion ratio. While, the increase of U-233 mole content in the core can increase the criticality and reduce the conversion ratio value. In addition, increasing LiF salt fraction on MSR FUJI-U3 fuel will make the criticality becomes less and conversion rate ratio becomes higher, while in the case of MSFR, more LiF salt fraction gives the increase of criticality condition and decrease the conversion ratio. After variations in fuel have been evaluated, an optimum composition is obtained based on criticality reactor condition and breeding condition. In case of MSR FUJI-U3, from the variation of U-233 as much as 0.22% -0.30% and variations in the LiF fraction of 51.76% -81.76% the best composition was 71.76% 〖LiF+16%BeF〗_2+〖12%ThF〗_4+〖0.24%UF〗_4 and 71.76% 〖LiF+16%BeF〗_2+〖11.98%ThF〗_4+〖0.26%UF〗_4. In addition, in case of MSFR type, based on the variations of U-233 as much as 2.0% -3.0% and variations in the fraction of LiF salts in the range of 73.5% -81.5%, the optimum composition is 77.5% LiF + 〖2.5%UF〗_4. In term of non-proliferation aspect, material barrier composition is employed that MSR FUJI-U3 and MSFR have a high level of intrinsic level proliferative-resistant which is based on material composition of U-232 and Pu-238 as material control for proliferation resistant. Keywords: Molten salt reactor, MSFR, criticality, conversion ratio, non-proliferation Topic: Nuclear Energy Expanded Applications

88 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-62] EXPERIMENTAL AND NUMERICAL STUDY OF FLOW ACCELERATED CORROSION IN A T-JUNCTION BY UTILIZING CFX AND 2D PIV

Ali Erfaninia(a,b,*) , Mohammadreza Nematollahi(a,b)

a) Shiraz University, School of Mechanical Engineering, Mollasadra St, 71348-51154, Shiraz, Iran. b) Shiraz University, Nuclear Safety Research Center, Nuclear Engineering Department, Shiraz, Iran. * [email protected]

Abstract

Flow Accelerated Corrosion (FAC) as one of the most important and degradation mechanisms of piping systems in nuclear and non-nuclear power plants results in thinning of the pipe wall and component shells from inside. One of the most susceptible areas to FAC is the T-junction, which is frequently used in power plant piping systems. The main mechanism of FAC occurrence results from the transfer of the corrosion products and iron ions to the bulk flow across the diffusion boundary layer. Using the mass transfer coefficient, the convective transfer of iron species from the metal surface into the bulk flow is described. In this study by developing a test facility and using 2D PIV, the flow field in a 90 degree straight T-junction with a 90 degree bend in the upstream is visualized. The flow field was simulated by using CFX and verified by PIV results. Using Chilton-Colburn analogy, hydrodynamic parameters such as the curvature ration of upstream elbow, the distance between the branch pipe and the upstream main pipe, the diameter of the branch pipe and the angle of connection of the branch pipe to the T- junction are considered and their effects on the changes of the mass transfer coefficient, changes of the magnitude of the mass transfer coefficient were investigated, separately.

Keywords: T-junction; 2D PIV; CFX; Mass Transfer Coefficient; FAC

Topic: Nuclear Energy Expanded Applications

89 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-63] 2D PIV STUDY OF FLOW ACCELERATED CORROSION DOWNSTREAM A TYPICAL INDUSTRIAL GATE VALVE

Abbas Sedghkerdar(a), Ali Erfaninia(a,b,*), Mohammadreza Nematollahi(a, b)

(a) Shiraz University, School of Mechanical Engineering, Nuclear Engineering Department, Shiraz, Iran. (b) Shiraz University, Nuclear Safety Research Center, Nuclear Engineering Department, Shiraz, Iran. * [email protected]

Abstract

A critical degradation mechanism of piping systems in either nuclear power plant is Flow Accelerated Corrosion (FAC), which causes wall thinning of pipes and component shells from inside. The valves which are frequently used in power plant piping systems are mostly subjected to FAC. A Gate vale in the feed water piping system of a typical nuclear power plant that it’s downstream had been damaged due to the FAC was considered as a reference. In this study the effect of the mean flow velocity and the opening percentage of gate valve on the flow field downstream the valve was investigated by using 2D PIV. The locus where flow turbulences happen downstream the gate valve was determined by 2D PIV. It is found that there is a very good coincidence between the locus in which the maximum wall thinning measured by Ultrasonic Technique (UT) and that the flow turbulences happen downstream the gate valve determined by 2D PIV. It is observed that the locus of flow turbulences and intense velocity gradients is in the upper part of the pipe and over the domain of 0.2

Keywords: Gate Valve; 2D PIV; FAC; Wall Thinning; UT

Topic: Nuclear Energy Expanded Applications

90 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-81] AG AND PD FISSION PRODUCT IMPLANTATION ON SIC LAYER IN TRISO FUEL PARTICLE OF HTGR USING SRIM/TRIM MONTE CARLO COMPUTER CODE

Mardiyanto Mangun Panitra1 and Abu Khalid Rivai1

1 Center For Science and Technology of Advanced Materials National Nuclear Energy Agency of Indonesia BATAN, Kawasan PUSPIPTEK Gd. 71, Tangerang Selatan, Banten, 15314, Indonesia Email: [email protected]

Abstract

Silicon Carbide (SiC) has some excellent characteristics such as wide band gap, high thermal conductivity, high electron mobility, and resistance to radiation effects. So that, SiC has been widely used for various applications including nuclear fuel system. SiC is used in TRISO (Tri-Structural Isotropic) coated fuel particle in HTGR (High Temperature Gas cooled Reactor). TRISO which consist of Inner Pyrolitic Carbon, SiC and Outer Pyrolitic Carbon is one of the safety systems feature of the reactor. However, one of the issues of the system is corrosion of SiC caused by silver (Ag) and palladium (Pd). Nevertheless, the detail mechanism of this corrosion phenomenon, such as the existence of Ag and Pd and how deep those two fission products penetrate into SiC layer, are still unknown. The objective of this study is to investigate the physical interaction of Ag and Pd with SiC coating layer of TRISO nuclear fuel particles. For this purpose, the physical effect of the penetration of the energetic Pd and Ag fission products into the SiC layer has been simulated using SRIM (Stopping and Range of Ions in Matter) /TRIM (TRansport of Ions in Matter) computer code with Monte Carlo method. In this simulation, various kinetic energies of Ag and Pd ion have been employed. The results showed the Ag/SiC and Pd/SiC Ion Ranges, Doses, Damage as the first-step evaluation to understand the corrosion phenomenon of the SiC-layer in the TRISO particles of HTGR.

Keywords: HTGR, SiC, Ag, Pd, SRIM/TRIM, corrosion

Topic: Nuclear Energy Expanded Applications

91 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-82] INVESTIGATION ON PHYSICAL INTERACTION OF AG AND PD FISSION PRODUCTS WITH ZRC LAYER IN TRISO FUEL PARTICLE OF HTGR USING SRIM/TRIM MONTE CARLO COMPUTER CODE

Mardiyanto Mangun Panitra1 and Abu Khalid Rivai1

1 Center For Science and Technology of Advanced Materials National Nuclear Energy Agency of Indonesia BATAN, Kawasan PUSPIPTEK Gd. 71, Tangerang Selatan, Banten, 15314, Indonesia Email: [email protected]

Abstract

High Temperature Gas cooled Reactors (HTGRs) as one type of Generation IV reactors that use TRISO (tri-structural isotropic) coated-fuel particles (CFP) for containment of radioactive fission products, which is produced from fission reaction of UO2 fuel. ZrC has been proposed to be the main barrier for containing fission products either as replacement of SiC layer or as an additional layer of the TRISO fuel particle to overcome the corrosion issue of SiC because of interaction with fission product of silver (Ag) and palladium (Pd). ZrC is an excellent material because it has good physical and nuclear properties i.e. high corrosion-resistant, low neutron capture cross section, good thermal shock resistance, thermal stability, etc. It is expected that ZrC to be a better barrier than SiC against the diffusion-attack of Ag and Pd. However, it depends on various factors such as chemical composition and interaction, microstructure, morphology, the presence of impurities, etc. Many attempts have been made to study the interaction phenomena of Ag and Pd with ZrC that cause the corrosion. Here, the physical interaction of those two fission products were studied and simulated with various kinetic energies using SRIM (Stopping and Range of Ions in Matter) /TRIM (TRansport of Ions in Matter) computer code with Monte Carlo method. The results obtained give detail information about the Ag/ZrC and Pd/ZrC Ion Ranges, Doses, Damage and corrosion attack. Furthermore, the depth and concentration of Ag and Pd fission product in ZrC have been observed, which is an important first step in understanding the corrosion phenomena of the ZrC-layer in the TRISO particles.

Keywords: HTGR, ZrC, SiC, Ag, Pd, SRIM/TRIM

Topic: Nuclear Energy Expanded Applications

92 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book

[ABS-83] MODELING OF DRAINAGE SYSTEM ON MODIFIED FREEZE VALVE IN A MOLTEN SALT REACTOR

Robi Dany Riupassa (a*), Abdul Waris (b), Khairul Basar (b), Novitrian (b), Yazid Bindar (c), Cici Wulandari (a)

a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia *[email protected] b) Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia c) Department of Chemical Engineering, Faculty of Industrial Engineering, Institut Teknologi Bandung, Indonesia

Abstract

Molten Salt Reactor (MSR) uses a freeze valve to overcome the reactor core temperature rise due to the decay heat after shutdown. The freeze valve functions as a blockage which will automatically open when the reactor core temperature rises. When the blockage opens, the fuel from the reactor core will flow to the emergency dump tank. In this study, a modification of the freeze valve design was developed from other researchers with a new type that would give more heat transfer modes. This new type consists of two parts, namely the head plug and neck plug and added heating rings. For further optimization, there is also an additional metal in the middle of the freeze valve to speed up the heat transfer process. The design specifications of this system will be tested for fuel in the form of FLiNaK (LiF-NaF-KF: 46.4-11.5-42 mole%). Heating rings, reactor vessels, and drainage pipes use Hastelloy-N material. Freeze valve uses Cesium Chloride (CsCl) material while heating metal is used copper (Cu). Computer modeling is based on computational fluid dynamics (CFD) to determine the melting behavior of a modified freeze valve.

Keywords: freeze valve; MSR; CFD

Topic: Nuclear Energy Expanded Applications

93 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

ICENES 2019 Program Book – Notes

ICENES2019 Holiday Inn Resort, Bali, Indonesia, October, 6-9th, 2019

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