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TR0700270

13™ INTERNATIONAL CONFERENCE ON EMERGING NUCLEAR ENERGY SYSTEMS

June 03 - 08, 2007 İstanbul, Türkiye ABSTRACTS

HOST ORGANIZATIONS Gazi University, Ankara Bahçeşehir University, Istanbul

MAJOR SPONSORS I S T C §§il iililSl M H T Permission is granted for single photocopies of single articles may be made for personal use to reproduce or distribute an individual abstract for educational purposes only. No commercial use or sale is permitted and on this purpose no part of this publication may be reproduced in any form, in an electronic retrieval system or otherwise, without the prior written permission of the publisher.

ISBN-978-975-01805-0-7

The copyright to the full work naturally remains with the editor or other current copyright holder. Any copyright questions regarding this publication work therefore should be addressed to the publisher. All other questions also relating to copyright and permissions should be addressed to:

Prof. Dr.-Ing. Sümer ŞAHİN Gazi Üniversitesi Teknik Eğitim Fakültesi Makina Bölümü Enerji Anabilim Dalı Teknikokullar-ANKARA 06503-TÜRKİYE Tel. + Fax: +90-312-212 43 04 E-mail: [email protected]

Printed in TÜRKİYE ICENES 2007

13»T1H" INTERNATIONAL CONFERENCE ON EMERGING NUCLEAR ENERGY SYSTEMS

ABSTRACTS

June 03 - 08, 2007 Istanbul, Türkiye

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

INTRODUCTION AND WELCOME to ICENES2007

The main objective of International Conference series on Emerging Nuclear Energy Systems (ICENES) is to provide an international scientific and technical forum for scientists, engineers, industry leaders, policy makers, decision makers and young 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.

One can observe worldwide that Nuclear Energy is in a true process of re-birth, and our success in this process will clearly depend on the emerging ideas we can offer for a sustainable future for Nuclear Energy. This conference is the perfect forum from which to offer relevant contributions towards this re-birth. In this century, the nuclear sciences representing cutting edge technology will play a major role in the formation of the intellectual assets of mankind taking us through to the century.

Previous ICENES conferences were held in Graz (Austria), Lausanne (Switzerland), Helsinki (Finland), Madrid (Spain), Karlsruhe (Germany), Monterey (USA), Chiba (Japan), Obninsk (Russia), Tel-Aviv (Israel), Petten (The Netherlands), Albuquerque (USA) and Brussels (Belgium). These conferences have enabled previously UNTHINKABLE IDEAS to be brainstormed into sound scientific concepts.

The first meeting in Graz in 1978 was initiated as a workshop with a small number of idealist scientists. The 1980 conference in Lausanne saw ICENES gain global respect and become an international ongoing annual event. Since then, the ICENES annual conference has made an important contribution to nuclear science and technology as a medium size International Conference.

It would be more accurate and constructive to consider that nuclear energy, and other newly emerging alternative energy technologies, are complementing and not competing with each other. For this reason, the new dimension of ICENES2007 has been to extend the forum, which also comprises innovative non-nuclear technologies such as energy and solar energy.

The Organization Committee is proud to announce that ICENES2007 had a record number of paper submissions in the ICENES conference series with 159 accepted papers from 35 countries, as shown in the table below. The main topics of these papers are fusion science and technology, fission reactors, accelerator driven systems, transmutation, laser in , shielding, nuclear reactions, hydrogen energy, solar energy, low energy physics and societal issues.

With ICENES2007, we cross the threshold from being of medium size to becoming a large International Conference in nuclear sciences. We hope the conference enhances the quality and quantity of international interactions, discussions and collaborations and widens the scientific horizon.

Sümer ŞAHİN 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

Table 1: Country listing of the presented papers at ICENES2007 based on the first authors

Country Papers 1. USA 25 2. Türkiye 23 3. Republic of Korea 17 4. Japan 13 5. Iran 11 6. Russia 9 7. Germany 8 8. Spain 6 9. Pakistan 5 10. France 4 11. Canada 3 12. Italy 3 13. Sweden 3 14. Belgium 2 15. Finland 2 16. Indonesia 2 17. Israel 2 18. Ukraine 2 19. Romania 2 20. Taiwan 2 21. Algeria 1 22. Austria 1 23. Bahrain 1 24. China 1 25. India 1 26. Netherlands 1 27. Nigeria 1 28. Poland 1 29. Portugal 1 30. Saudi Arabia 1 31. Slovenia 1 32. Switzerland 1 33. UK 1 34. Uzbekistan 1 35. Venezuela 1 36. IAEA 1 Total 160 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ORGANIZATION COMMITTEE

Prof. Dr. Sümer ŞAHİN Conference Chairman Prof. Dr. Jose M. MARTINEZ-VAL General Chairman Prof. Dr. Şenay YALÇIN Conference Co-Chairman Assoc. Prof. Dr. H. Mehmet ŞAHİN Technical Program Chairman Assist. Prof. Dr. Adem ACIR Scientific Secretary Erkut TANRISEVEN Congress Organizer

International Scientific Committee C. RUBBIA, Italy, Honorary K. Y. SUH, Republic of Korea N. DIAZ, USA, Honorary L. EL-GUEBALY, USA T. N. VEZİROĞLU, Türkiye, Honorary L. TEPPER, Israel A. MILLER, Canada L. TOCHENY, Russia A. SHENOY, USA L. VANHOENACKER, Belgium A. H. M. VERKOOIJEN, The Netherlands M. ABDOU, USA B. CARLUEC, France M. S. EL-GENK, USA E. CHENG, USA M. DECRETON, Belgium E. GREENSPAN, USA P. VAZ, Portugal F. SEFIDVASH, Brazil R. KARAM, USA H. Ait ABDERRAHIM, Belgium R. MOIR, USA H. SEKIMOTO, Japan R. SALOMAA, Finland H. TAKAHASHI, USA S. ANGHAIE, USA İ. DİNÇER, Canada S. ELIEZER, Israel J. DEKEYSER, Belgium T. KAMASH, USA J. A. RUBIO, Spain T. MEHLHORN, USA J.C. KUIJPER, The Netherlands W. GUDOWSKI, Sweden K. FURUKAWA, Japan X. VITART, France K. SCHÖPF, Austria Y. RONEN, Israel K. SCHULTZ, USA

National Scientific Committee O. ÇAKIROĞLU N. K. ARAS E. ARIK N. ASLAN S. SULTANSOY V. ALTIN K. ALTINIŞIK O. ALNIAK F. KADIRGAN M.ÜBEYLİ H. SAYGIN O. İPEK H. YAPICI K. YILDIZ T. ALTINOK N. ŞAHIN E. TANKER E. TEL B. ŞARER 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

HOST ORGANIZATIONS

• Gazi University, Ankara • Bahçeşehir University, Istanbul

MAJOR SPONSORS

• Ministry of Culture and Tourism • Turkish Atomic Energy Authority (TAEK) • Turkish Scientific and Technical Research Council (TÜBİTAK) • International Centre for Hydrogen Energy Technologies of United Nations Industrial Development Organization (UNIDO- ICHET) • International Science and Technology Center (ISTC) 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

13TH INTERNATIONAL CONFERENCE ON EMERGING NUCLEAR ENERGY SYSTEMS

PROGRAMME

SUNDAY, 03 JUNE 2007

REGISTRATION

(15:30-20:30. World Park Hotel)

MONDAY, 04 JUNE 2007

REGISTRATION (8:00-9:00 World Park Hotel) OPENING SESSION (9:30-10.00) Prof. Dr. Sümer Şahin; Conference Chairman Prof. Dr. Jose M. Martinez-Val, General Chairman Prof. Dr. Kadri Yamaç, President, Gazi University Okay Çakıroğlu, Chairman, Turkish Atomic Energy Authority

COFFEE BREAK (10:00-10:30)

SESSION 1 PLENARY SESSION (10:30-12:20)

Chairman: Guillemo Velarde

KEYNOTE SPEECH: OVERVIEW OF PRINCIPLES AND CHALLENGES OF FUSION NUCLEAR TECHNOLOGY Mohamed Abdou, Los Angeles, USA

A ROAD MAP FOR THE REALIZATION OF GLOBAL-SCALE BREEDING FUEL CYCLE BY SINGLE MOLTEN-FLUORIDE FLOW Kazuo Furukawa, et al., Kanagawa, Japan

NIF: A PATH TO FUSION ENERGY Edward Moses, Livermore, USA LUNCH (12:20-13:30) 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SESSION 2 PLENARY SESSION (14:00-16:00)

Chairman: Ralph W. Moir

SPACE CONCEPTS FOR AVOIDANCE OF A SINGLE POINT FAILURE M. S. El-Genk, Albuquerque, USA

IRRADIATION FACILITIES FOR MATERIALS RESEARCH: IFMIF AND SMALL SCALE INSTALLATIONS J. M. Perlado, Madrid, Spain M. Victoria, Madrid, Spain

SUSTAINABLE ENERGY SYSTEMS AND THE EURATOM RESEARCH PROGRAMME S. Webster, Brussels, BE G. Van Goethem, Brussels, BE

R&D ACTIVITIES OF THE JOINT RESEARCH CENTRE (JRC) AND ITS INVOLVEMENT IN THE DEVELOPMENT OF FUTURE NUCLEAR ENERGY SYSTEMS Roland Schenkel, Germany

COFFEE BREAK (16:00-16:20)

SESSION 3

SESSION 3A THE FUTURE OF (16:20-18:00)

Chairman: Borut Smodis

INNOVATIVE FISSION REACTORS FOR THIS CENTURY Emilio Minguez, Madrid, Spain

NUCLEAR POWER IN SPACE Samim Anghaie, Florida, USA

THE FIXED BED NUCLEAR REACTOR CONCEPT Sümer Şahin, Ankara, Türkiye Farhang Sefidvash, Porto Alegre, Brazil

CHALLENGES OF FOR THE SUSTAINABLE ROLE IN KOREAN ENERGY POLICY Young Eal Lee, Korea

REACTOR PHYSICS IDEAS TO DESIGN NOVEL REACTORS WITH FASTER FISSILE GROWTH V. Jagannathan et. al., Mumbai, India 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SESSION 3B ACCELERATOR DRIVEN SYSTEMS I (16:20-18:00)

Chairman: Werner Maschek

DEVELOPMENT OF THE HEAT SINK STRUCTURE OF A BEAM DUMP FOR THE PROTON ACCELERATOR Wan Young Maeng et. al., Korea

SUBCRITICALITY DETERMINATION IN ADS: THE YALINA-BOOSTER EXPERIMENTS Carl-Magnus Persson, Stockholm, Sweden Waclaw Gudowski, Stockholm, Sweden Andrei Fokau et. al., Belarus

STUDY OF FISSION CROSS SECTIONS INDUCED BY NUCLEONS AND PIONS USING THE CASCADE-EXCITON MODEL CEM95 Zafar Yasin, Islamabad, Pakistan M. ikram Shahzad, Islamabad, Pakistan

MONTE CARLO STUDIES IN ACCELERATOR-DRIVEN SYSTEMS FOR TRANSMUTATION OF HIGH-LEVEL NUCLEAR WASTE B. Şarer, Ankara, Türkiye M. E. Korkmaz, Ankara, Türkiye M. Günay, Malatya, Türkiye A. Aydm, Kırıkkale, Türkiye

CALCULATIONS OF THE MAIN FREE PATH ON EMISSION CROSS-SECTION FOR SPALLATION REACTION OF TARGET AND FUEL NUCLEI E. Tel, Gazi Univ., Türkiye H. F. Kışoğlu, Adana, Türkiye A. Kayış Topaksu, Adana, Türkiye A. Aydın, İsparta, Türkiye A. Kaplan, İsparta, Türkiye

SESSION 3C ALTERNATIVE ENERGY (16:20-18:00)

Chairman: Kazuhiko K. Kudo

ARE THE LESSONS OF THE TECHNO-ECONOMIC STUDIES ON THE SULPHUR-IODINE CYCLE APPLICABLE TO THE OTHER CYCLES? F. Werkoff, France C. Mansilla, France

PRODUCTION METHOD OF HYDROGEN BY JET PLASMA PROCESS IN HYDRO MACHINERY Farzan Amini, Tehran, Iran

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ANALYSIS OF AN INNOVATIVE SOLAR WATER DESALINATION SYSTEM USING INDUCED VACUUM Teoman Ayhan, Bahrain Husain Al-Madani, Bahrain

EXPERIMENTAL INVESTIGATION TWO PHASE FLOW IN DIRECT METHANOL FUEL CELLS Mahmut D. Mat, Niğde, Türkiye Yüksel Kaplan, Niğde, Türkiye Selahattin Çelik, Niğde, Türkiye Aytekin Öztorul, Niğde, Türkiye

POTENTIAL OF HYDROGEN PRODUCTION FROM WIND ENERGY IN PAKISTAN Mohammad Aslam Uqaili, Pakistan Khanji Harijan, Pakistan Mujeebuddin Memon, Pakistan

WELCOME RECEPTION AT BAHÇEŞEHİR UNIVERSITY (19:30) (Departure from World Park Hotel at 18:45)

TUESDAY, 05 JUNE 2007

SESSION 4

SESSION 4A FUSION I (08:40-10:20)

Chairman: Jose M. Perlado

NUCLEAR CHALLENGES AND PROGRESS IN DESIGNING POWER PLANTS Laila El-Guebaly, Madison, USA

SELF-ORGANIZED IGNITION OF A PLASMA Klaus Schoepf, Innsbruck, Austria

THE STUDY OF INERTIAL FUSION ENERGY PROBLEM VIA THE EQUATION OF STATE Shalom Eliezer, Yavne, Israel J.M. Martinez Val, Madrid, Spain M. Murakami, Osaka, Japan

TRIANGULARITY EFFECTS ON THE COLLISIONAL DIFFUSION FOR ELLIPTIC TOKAMAK PLASMA P. Martin, Venezuela E. Castro, Venezuela

INNOVATIVE THREE-DIMENSIONAL NEUTRONICS ANALYSES DIRECTLY COUPLED WITH CAD MODELS OF GEOMETRICALLY COMPLEX FUSION SYSTEMS 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

Mohamed Sawan, Madison, USA P. Wilson, Argonne, USA T. Tautges, Madison, USA L. El-Guebaly, Madison, USA D. Henderson, Madison, USA G. Sviatoslavsky, Madison, USA T. Bohm, Madison, USA B. Kiedrowski, Madison, USA A. Ibrahim, Madison, USA B. Smith, Madison, USA R. Slaybaugh, Madison, USA

SESSION 4B FISSION REACTORS I (08:40-10:20)

Chairman: Emilio Minguez

NAVAL APPLICATION OF BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM Nam H. Kim, Seoul, Korea Tae W. Kim, Seoul, Korea Hyoung M. Son, Seoul, Korea Kune Y. Suh, Seoul, Korea

ENHANCING VVER ANNULAR PROLIFERATION RESISTANCE FUEL WITH MINOR G. S. Chang, Idaho, USA

PRELIMINARY DESIGN OF SMART FUEL Yonghwan Kim, Duck-Jin Dong Dae-Jeon City, Korea Dongguen Ha, Duck-Jin Dong Dae-Jeon City, Korea Sungkyu Park, Duck-Jin Dong Dae-Jeon City, Korea Keeil Nahm, Duck-Jin Dong Dae-Jeon City, Korea Kyuseok Lee, Duck-Jin Dong Dae-Jeon City, Korea Jungha Kim, Duck-Jin Dong Dae-Jeon City, Korea

SOLITARY BURN-UP WAVE SOLUTION IN MULTI-GROUP DIFFUSION-BURNUP COUPLED SYSTEM X.-N. Chen, FZK, Germany W. Maschek, FZK, Germany

WHOLE CORE TRANSPORT CALCULATION FOR THE VHTR HEXAGONAL CORE Jin-Young Cho, Daejeon, Korea Kang-Seog Kim, Daejeon, Korea Chung-Chan Lee, Daejeon, Korea Han-Gyu Joo, Seoul, Korea 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SESSION 4C SOLAR ENERGY (08:40-10:20)

Chairman: Mohammadreza Nematollahi

MICROSCOPIC MORPHOLOGICAL COMPENSATION FOR PHASE-SEPARATED COMPOSITE FILM THICKNESS Chia -Fu Chang, Yilan City, Taiwan Yi-Ci Chan, Yilan City, Taiwan Zou-ni Wan, Yilan City, Taiwan

A NOVEL GUI MODELED FUZZY LOGIC CONTROLLER FOR A SOLAR POWERED ENERGY UTILIZATION SCHEME I. H. Altas, Trabzon, Türkiye A.M. Sharaf, New Brunswick, Canada

DESIGN AND PERFORMANCE OF GAN BETAVOLTAIC DEVICE Hyun-Kyu Jung, Daejon, Korea Nam-Ho Lee, Daejon, Korea Sang-Kwon Lee, Jeonju, Korea

SOLAR CONTROL ON IRRADIATED TA2O5 THIN FILM Nilgün Doğan Baydoğan, Istanbul, Türkiye Esra Özkan Zayim, İstanbul, Türkiye

CONCEPTUAL DESIGN OF GAN BETAVOLTAIC BATTERY USING IN CARDIAC PACEMAKER M. Mohamadian, Tehran, Iran S.A.H. Feghhi, Tehran, Iran H. Afarideh, Tehran, Iran

COFFEE BREAK (10:20-10:40)

SESSION 5

SESSION 5A ACCELERATOR DRIVEN SYSTEMS II (10:40-12:20)

Chairman: Edward Moses

R&D ACTIVITIES AROUND THE EUROTRANS ACCELERATOR FOR ADS APPLICATIONS Jean-Luc Biarrotte, France Alex C. Mueller, France

NEUTRONICS AND SHIELDING ISSUES OF ADS Hamid A. Abderrahim, Belgium T. Aoust, Belgium

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E. Gonzalez, Spain W. Haeck, Belgium E. Malambu, Belgium J. M. Martinez-Val, Spain Y. Romanets, ITN, Portugal G. Van den Eynde, Belgium P. Vaz, Portugal C. Vicente, Spain THERMAL FEATURES OF SPALLATION WINDOW TARGETS J. M. Martinez-Val, Madrid, Spain F. Sordo, Madrid, Spain P. T. Leon, Madrid, Spain

CERMET FUEL BEHAVIOUR AND PROPERTIES IN ADS REACTORS D. Haas, Germany A. Fernandez, Germany D. Staicu, Germany J. Somers, Germany W. Maschek, Germany X. Chen, Germany

ASSESSMENT OF THE TRANSMUTATION CAPABILITY OF AN ACCELERATOR DRIVEN SYSTEM COOLED BY LEAD BISMUTH EUTECTIC ALLOY F. Bianchi, Bologna, Italy V. Peluso, Bologna, Italy R. Calabrese, Bologna, Italy X. Chen, Karlsruhe, Germany W. Maschek, Karlsruhe, Germany SESSION 5B FISSION REACTORS II (10:40-12:20)

Chairman: Masaki Ozawa

REACTOR PHYSICS SIMULATIONS WITH COUPLED MONTE CARLO CALCULATION AND COMPUTATIONAL FLUID DYNAMICS Volkan Seker, Purdue, USA Justin W. Thomas, Purdue, USA Thomas J. Downar, Purdue, USA

FORMULATION OF DETECTOR RESPONSE FUNCTION TO CALCULATE THE POWER DENSITY PROFILES USING IN-CORE NEUTRON DETECTORS ASM Sabbir Ahmed, Germany J. K. Shultis, USA Joerg Peter, Germany Wolfrad Semmler, Germany

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SCENARIO ANALYSIS FOR TRANSURANIC TRANSMUTATION BY USING FAST REACTORS Chang Joon Jeong, Daejeon, Korea CALCULATION OF NEUTRON IMPORTANCE FUNCTION IN FISSIONABLE ASSEMBLIES USING MONTE CARLO METHOD S. A. H. Feghhi, Tehran, Iran M. Shahriari, Tehran, Iran H. Afarideh, Tehran, Iran

SENSORY SYSTEMS FOR A CONTROL ROD POSITION USING REED SWITCHES FOR THE INTEGRAL REACTOR Je-Yong Yu, Daejon, Korea Suhn Choi, Daejon, Korea Ji-Ho Kim, Daejon, Korea Doo-Jeong Lee, Daejon, Korea SESSION 5C ENERGY TECHNOLOGIES (10:40-12:20)

Chairman: Pedro Vaz

AMBIENT TEMPERATURE EFFECTS ON GAS TURBINE POWER PLANT: A CASE STUDY IN IRAN Mofid Gorji, Babol, Iran Fama Fouladi, Babol, Iran

ENERGY CONSERVATION THROUGH THE IMPLEMENTATION OF CO-GENERATION & GRID INTERCONNECTION M. A. Dashash, Saudi Aramco

A NOVEL CO2 SEQUESTRATION SYSTEM FOR ENVIRONMENTALLY PRODUCING HYDROGEN FROM FOSSIL-FUELS William Eucker IV, Maryland, USA MECHANICAL ALLOYING OF MG-CO-NI POWDER FOR HYDROGEN STORAGE Hadi Suwarno, Banten, Indonesia Andon insani, Banten, Indonesia Johny Wahyudi, West Java, Indonesia Eddy S. Siradj, West Java, Indonesia Bambang Herutomo, Banten, Indonesia

CONTRIBUTION OF WIND ENERGY TO FUTURE ELECTRICITY REQUIREMENTS OF PAKISTAN Khanji Harijan, Pakistan Muhammad Aslam Uqaili, Pakistan Mujeebuddin Memon, Pakistan

LUNCH (12:20-13:30)

12 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SESSION 6 PLENARY SESSION (14:00-16:00)

Chairman: Shalom Eliezer

RECOMMENDATIONS FOR A RESTART OF DEVELOPMENT Ralph W. Moir, Livermore, USA

INNOVATIVE ICF SCHEME - IMPACT FAST IGNITION M. Murakami, Osaka, Japan H. Nagatomo, Osaka, Japan T. Sakaiya, et. al., Osaka, Japan

PURSUING NUCLEAR ENERGY WITH NO NUCLEAR CONTAMINATION —FROM REACTOR TO DEUTERON FLUX REACTOR— Xing Z. Li, Beijing, China Qing M. Wie, Beijing, China Bin Liu, Beijing, China Xie G. Zhu, Beijing, China Shao L. Ren, Beijing, China

NUCLEAR REACTOR DEVELOPMENT IN KOREA: IT'S HISTORY AND STATUS Jongsik Cheong, Daejeon, Korea Insik Kim, Daejeon, Korea Dong-Su Kim, Daejeon, Korea

COFFEE BREAK (16:00-16:20)

SESSION 7

SESSION 7A FISSION REACTORS III (16:20-18:00)

Chairman: Samim Anghaie

TRANSMUTATION OF MINOR ACTINIDES IN A CANDU THORIUM BURNER Sümer Şahin, Ankara, Türkiye Şenay Yalçın, İstanbul, Türkiye Hacı Mehmet Şahin, Ankara, Türkiye Adem Acır, Ankara, Türkiye Kadir Yıldız, Akasaray, Türkiye Necmettin Şahin, Akasaray, Türkiye Taner Altmok, Ankara, Türkiye Mahmut Alkan, Niğde, Türkiye

13 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

4+D DIGITAL ENGINEERING FOR ADVANCED NUCLEAR ENERGY SYSTEMS Seo G. Jeong, Seoul, Korea Seung K. Nam, Seoul, Korea Kune Y. Suh, Seoul, Korea

FEASIBILITY STUDY OF SELF SUSTAINING CAPABILITY ON WATER COOLED THORIUM REACTORS FOR DIFFERENT POWER REACTORS Sidik Permana, Tokyo, Japan Naoyuki Takaki, Tokyo, Japan Hiroshi Sekimoto, Tokyo, Japan

GENERATION IV NUCLEAR PLANT DESIGN STRATEGIES Vural Altın, Ankara, Türkiye

TYPICAL STEAM GENERATOR TUBES RUPTURE (SGTR) EFFECT ON THERMO HYDRAULIC PARAMETERS OF VVER-1000 PRIMARY LOOP Abbas Zare, Shiraz, Iran Mohammadreza Nematollahi, Shiraz, Iran Kamal Hadad, Khosro Jafarpoor, Shiraz, Iran Khosro Jafarpoor, Shiraz, Iran Masood Aminmozaffari, Shiraz, Iran

SESSION 7B HYDROGEN ENERGY (16:20-18:00)

Chairman: James C. Kuijper

SOLAR PUMPED LASER AND ITS APPLICATION TO HYDROGEN PRODUCTION K.Imasaki, Osaka, Japan T. Saiki, Osaka, Japan D.Li, Osaka, Japan S.Motokosi, Osaka, Japan M.Nakatsuka, Osaka, Japan

NUMERICAL STUDY OF HYDROGEN ABSORPTION IN A LM-NI5 HYDRIDE REACTOR Kemal Altınışık, Konya, Türkiye Muhittin Tekin, Konya, Türkiye Mahmut D. Mat, Niğde, Türkiye Alper Altınışık, İstanbul, Türkiye T. Nejat Veziroğlu, Miami, U.S.A.

21ST CENTURY'S ENERGY: HYDROGEN ENERGY SYSTEM T. Nejat Veziroğlu, University of Miami, USA

TECHNOLOGIES FOR HYDROGEN PRODUCTION BASED ON DIRECT CONTACT OF GASEOUS HYDROCARBONS AND EVAPORATED WATER WITH MOLTEN PB OR PB- A. V. Gulevich, Obninsk, Russia P. N. Martynov, Obninsk, Russia

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V. A. Gulevsky, Obninsk, Russia V. V. Ulyanov, Obninsk, Russia

THERMAL DESIGN, TECHNICAL ECONOMICAL AND ENVIRONMENTAL ANALYSES OF A HYDROGEN FIRED MULTI-OBJECTIVE COGENERATION SYSTEM Ali Durmaz, Ankara, Türkiye M. Zeki Yılmazoglu, Ankara, Türkiye Ayşegül Pasaoğlu, Ankara, Türkiye

SESSION 7C FUSION & PHYSICS (16:20-18:00)

Chairman: Yasuyuki Nakao

RADIATION-PHYSICAL PROCESSES AT NUCLEAR-TRANSMUTATION OF SILICON DOPED BY PALLADIUM Sh. Makhkamov, Tashkent, Uzbekistan N. A. Tursunov, Tashkent, Uzbekistan A. R. Sattiev, Tashkent, Uzbekistan

LASER-DRIVEN HIGH-ENERGY IONS AND THEIR APPLICATION TO INERTIAL CONFINEMENT FUSION M. Borghesi, The Queen's University of Belfast (UK) PLASMA FOCUS SYSTEM; DESIGN, CONSTRUCTION AND EXPERIMENTS A. Alaçakır, Ankara, Türkiye Y. Akgün, Ankara, Türkiye A. S. Bölükdemir, Ankara, Türkiye A. Elmalı, Istanbul, Türkiye H. Karadeniz, Ankara, Türkiye T. Öncü, Ankara, Türkiye E. Recepoğlu, Ankara, Türkiye İ. T. Çakır, Ankara, Türkiye Ö. Yeşiltaş, Ankara, Türkiye S. Zararsız, Ankara, Türkiye

NEW DISCOVERY: QUANTIZATION OF ATOMIC AND NUCLEAR REST MASS DIFFERENCES AND SELF-ORGANIZATION OF ATOMS AND NUCLEI F.A. Gareev, Dubna, Russia I.E. Zhidkova, Dubna, Russia

THE ALBEDO PROBLEM FOR PURE-QUADRATIC SCATTERING D. Türeci, Ankara, Türkiye R. G. Türeci, Kırıkkale, Türkiye

DINNER AT BALTALİMANI, Sponsored by Turkish Atomic Energy Authority TAEK (Departure from World Park Hotel at 18:30)

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WEDNESDAY, 06 JUNE 2007

SESSION 8

SESSION 8A FISSION REACTORS IV (08:40-10:20)

Chairman: Tawfik Al-Kusayer

NOBLE GAS, BINARY MIXTURES FOR COMMERCIAL GAS-COOLED REACTOR SYSTEMS Mohamed S. El-Genk, Albuquerque, USA Jean-Michel Tournier, Albuquerque, USA

ACR-1000: OPERATOR-BASED DEVELOPMENT B.Shalaby, Atomic Energy of Canada Limited, Canada A.Alizadeh, Atomic Energy of Canada Limited, Canada

PLUTONIUM AND MINOR ACTINIDES MANAGEMENT IN THERMAL HIGH-TEMPERATURE REACTORS - THE EU FP6 PROJECT PUMA J. C. Kuijper et al., The Netherlands

DESIGN STUDY ON SMALL CANDLE REACTOR Hiroshi Sekimoto, Tokyo, Japan Mingyu Yan, Tokyo, Japan

MONTE CARLO BENCHMARK CALCULATIONS FOR 400MWTH PBMR CORE Hong-Chul KIM, Seoul, Korea Soon Young KIM, Seoul, Korea Jong Kyung KIM, Seoul, Korea Jae Man NOH, Seoul, Korea SESSION 8B LASERS & NUCLEAR REACTORS (08:40-10:20)

Chairman: Klaus Schoepf

PULSE REACTOR SYSTEM FOR NUCLEAR PUMPED LASER USING LOW ENRICHED Toru Obara, Tokyo, Japan Hiroki Takezawa, Tokyo, Japan

DESIGN CHARACTERISTICS OF REGIONAL ENERGY REACTOR, REX-10 J. W. Kim, Seoul, Korea H. K. Ahn, Seoul, Korea H. M. Joo , Hanyang, Korea B. I. Jang, Hanyang, Korea M. H Kim, Kyunghee, Korea

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H. G. Joo, Seoul, Korea G. C. Park, Seoul, Korea

APPLICATIONS OF SUPER-HIGH INTENSITY LASERS IN Rainer Salomaa, Helsinki, Finland Antti Hakola, Helsinki, Finland Marko Santala, Helsinki, Finland

A STUDY ON PHYSICAL CHARACTERISTICS OF SUPERCRITICAL LIGHT-WATER REACTOR LOADED WITH (233U-TH-238U) OXIDE FUEL E.G.Kulikov, Moscow, Russia N.Shmelev, Moscow, Russia G.G.Kulikov, Moscow, Russia V.A.Apse, Moscow, Russia

NEUTRONICS OF A SALT COOLED - VERY HIGH TEMPERATURE REACTOR J. Zakova, Sweden

SESSION 8C EFFECTS (08:40-10:20)

Chairman: Khalid Aleissa

THE POSSIBILITY OF INTRODUCTION THE TEMPORAL SCALE HUBBLE IN DYNAMICAL REACTION AU+AU AT 200 AGEV STUDIED IN BRAHMS EXPERIMENT (BROOKHAVEN - USA) C. Besliu, Bucharest, Romania A. Danu, Bucharest, Romania Al. Jipa, Bucharest, Romania I.S. Zgura, Bucharest, Romania

NEUTRONIC LIMITS IN VARIOUS TARGET MEDIUMS DRIVEN BY A PROTON BEAM OF 1 GEV ENERGY Gamze Genç, Kayseri, Türkiye Nesrin Demir, Kayseri, Türkiye Hüseyin Yapıcı, Kayseri, Türkiye

BURNUP STUDIES OF THE SUBCRITICAL FUSION-DRIVEN IN-ZINERATOR Carl-Magnus Persson, Stockholm, Sweden Waclaw Gudowski, Stockholm, Sweden Francesco Venneri, California, USA

TRANSMUTATION OF HIGH LEVEL WASTES IN A FUSION-DRIVEN TRANSMUTER (FDT) Nesrin Demir, Kayseri, Türkiye Gamze Genç, Kayseri, Türkiye Hüseyin Yapıcı, Kayseri, Türkiye

COFFEE BREAK (10:20-10:40)

17 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SESSION 9 PLENARY SESSION (10:40-12:20)

Chairman: Ali Alizadeh

ADVANCED AND SUSTAINABLE FUEL CYCLES FOR INNOVATIVE REACTOR SYSTEMS J.-P. Glatz, Karlsruhe, Germany R. Malmbeck, Karlsruhe, Germany D. Serrano-Purroy, Karlsruhe, Germany P. Soucek, Karlsruhe, Germany T. Inoue, Tokyo, Japan K. Uozumi, Tokyo, Japan

GAMMA RAY BEAM TRANSMUTATION K. Imasaki, Osaka, Japan D. Li, Osaka, Japan S. Miyamoto, Osaka, Japan S. Amano, Osaka, Japan T. Motizuki, Osaka, Japan

RESULTS OF THE IAEA CRP ON 'STUDIES OF ADVANCED REACTOR TECHNOLOGY OPTIONS FOR EFFECTIVE INCINERATION OF ' W. Maschek et. al., Karlsruhe, Germany

THE PROMISES AND CHALLENGES OF FUTURE REACTOR SYSTEM DEVELOPMENTS Si-Hwan Kim, Daejon, Korea Moon Hee Chang, Daejon, Korea Hyun-Jun Kim, Daejon, Korea

LUNCH (12:20-13:30) SOCIAL EVENT: ISTANBUL CLASSICS (14:00) (Departure from World Park Hotel)

THURSDAY, 07 JUNE 2007

SESSION 10

SESSION 10A FISSION REACTORS V (08:40-10:20)

Chairman: Rainer Salomaa STATUS AND OVERVIEW OF INPRO Masanao Moriwaki, Vienna, Austria

18 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ADVANCED ORIENT CYCLE, FOR STRATEGIC SEPARATION, TRANSMUTATION AND UTILIZATION OF NUCLIDES IN THE CYCLE Masaki Ozawa, Tokyo, Japan Reiko Fujita, Tokyo, Japan Shinichi Koyama, Tokyo, Japan Tatsuya Suzuki, Tokyo, Japan Yasuhiko Fujii, Tokyo, Japan

TECHNO-ECONOMIC STUDY OF HYDROGEN PRODUCTION BY HIGH TEMPERATURE ELECTROLYSIS COUPLED WITH AN EPR - WATER STEAM PRODUCTION AND COUPLING POSSIBILITIES R. Rivera-Tinoco, Paris, France C. Mansilla, Gif sur Yvette CEDEX, France F. Werkoff, Gif sur Yvette CEDEX, France C. Bouallou, Paris, France

HYBRID NUCLEAR CYCLES FOR NUCLEAR FISSION SUSTAINABILITY Mireia Piera, Madrid-Spain Jose M. Martinez-Val, Madrid-Spain

POLITICS OF NUCLEAR POWER AND FUEL CYCLE Rizwan-uddin, Urbana-Champaign, USA SESSION 10B NEW PERSPECTIVES (08:40-10:20)

Chairman: Didier Haas

ANTIPROTON-INDUCED FISSION FOR SPACE POWER AND PROPULSION APPLICATIONS Terry Kammash, Michigan, USA

PRELIMINARY NEUTRONIC DESIGN OF SPOCK REACTOR: A NUCLEAR SYSTEM FOR SPACE POWER GENERATION N. Burgio, Rome - Italy M. Cumo, Rome - Italy A. Fasano, Rome - Italy M. Frullini, Rome - Italy A. Santagata, Rome - Italy

PREMISES FOR USE OF FUSION SYSTEMS FOR WASTE INCINERATION Stefan Taczanowski

AN ADVANCED ENERGY SYSTEM WITH NUCLEAR REACTORS AS AN ENERGY SOURCE Yasuyoshi Kato, Tokyo, Japan Takao Ishizuka, Tokyo, Japan K. Nikitin, Tokyo, Japan

19 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

THE GDT-BASED FUSION NEUTRON SOURCE AS DRIVER OF A MINOR ACTINIDES BURNER K. Noack, Dresden, Germany A. Rogov, Dresden, Germany A. A. Ivanov, Novosibirsk, Russia E. P. Kruglyakov, Novosibirsk, Russia Yu A. Tsidulko, Novosibirsk, Russia SESSION IOC REACTOR PHYSICS (08:40-10:20)

Chairman: Jasmina Vujic

MONTE CARLO CALCULATION FOR DIFFERENT ENRICHMENT LITHIUM MODERATOR IN A HYBRID REACTOR Hacı Mehmet Şahin, Ankara, Türkiye Şenay Yalçın, İstanbul, Türkiye Taner Altınok, Ankara, Türkiye Adem Acır, Ankara, Türkiye

LINEAR PULSE MOTOR TYPE CONTROL ELEMENT DRIVE MECHANISM FOR THE INTEGRAL REACTOR Je-Yong Yu, Daejon, Korea Suhn Choi, Daejon, Korea Ji-Ho Kim, Daejon, Korea Doo-Jeong Lee, Daejon, Korea

NEUTRONIC AND THERMAL HYDRAULIC ASSESSMENT OF FAST REACTOR COOLING BY WATER OF SUPER CRITICAL PARAMETERS Yu.D.Baranaev, Obninsk, Russia A.P.Glebov, Obninsk, Russia V.F.Ukraintsev, Obninsk, Russia V.V.Kolesov, Obninsk, Russia

SIMULATING A PARTIAL LOCA IN A NARROW CHANNEL USING THE DSNP SIMULATION SYSTEM D. Saphier, Yavne, Israel COFFEE BREAK (10:20-10:40)

SESSION 11

PLENARY SESSION (10:40-12:20)

Chairman: Mohamed S. El-Genk

EURATOM STRATEGY TOWARDS FUSION ENERGY Carlos Varandas, Portugal

20 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

FAST IGNITION STUDIES AT OSAKA UNIVERSITY Kazuo A. Tanakaand FI Project Team, Osaka, Japan

EXERGY ANALYSIS OF A SYSTEM USING A CHEMICAL HEAT PUMP TO LINK A SUPERCRITICAL WATER-COOLED NUCLEAR REACTOR AND A THERMO CHEMICAL WATER SPLITTING CYCLE Mikhail Granovskii, Ontario, Canada Ibrahim Dinçer, Ontario, Canada Marc A. Rosen, Ontario, Canada Igor Pioro, Ontario, Canada

LASER ENHANCED RADIOACTIVE DECAY AND SELECTIVE TRANSMUTATION OF NUCLEAR WASTE Rainer Salomaa, Helsinki, Finland Pertti Aarnio, Helsinki, Finland Jarmo Ala-Heikkila, Helsinki, Finland Antti Hakola, Helsinki, Finland Marko Santala, Helsinki, Finland

LUNCH (12:20-13:30)

SESSION 12

SESSION 12A NUCLEAR FUTURE (14:00-15:20)

Chairman: Hiroshi Sekimoto

RECENT PROGRESS IN STOCHASTIC TRANSPORT THEORY A. Ziya Akçasu, Michigan, USA

INNOVATIONS ON NUCLEAR ENERGY - WHAT CAN A SMALL COUNTRY CONTRIBUTE? Borut Smodis, Ljubljana, Slovenia Milan Cercek, Ljubljana, Slovenia

NUCLEAR POWER AS A NECESSARY OPTION, ALBEIT AN INSUFFICIENT ONE Vural Altın, Ankara, Türkiye

TOWARDS A NEW WORLD: THE CONTRIBUTIONS OF NUCLEAR ENERGY TO A SUSTAINABLE FUTURE R. B. Duffey, Chalk River, ON, Canada A. I. Miller, Chalk River, ON, Canada P. J. Fehrenbach, Chalk River, ON, Canada S. Kuran, Chalk River, ON, Canada D. Tregunno, Chalk River, ON, Canada S. Suppiah, Chalk River, ON, Canada

21 13" International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SESSION 12B RADIATION EFFECTS (14:00-15:20)

Chairman: Romney B. Duffey

RADIATION SHIELDING PERFORMANCE OF SOME CONCRETE İ. Akkurt, İsparta, Türkiye H. Akyıldırım, İsparta, Türkiye B. Mavi, İsparta, Türkiye S. Kılınçarslan, İsparta, Türkiye C. Başyiğit, İsparta, Türkiye

TRANSMUTATION OF TECHNETIUM INTO STABLE RUTHENIUM IN HIGH FLUX CONCEPTUAL RESEARCH REACTOR N. Amrani, Setif, Algeria A. Boucenna, Setif, Algeria

ENCAPSULATING OF HIGH-LEVEL RADIOACTIVE WASTE WITH USE OF PYROCARBON AND SILICON CARBIDE COATINGS A. Chernikov, Moscow, Russia

DETERMINATION OF SHIELDING PARAMETERS FOR DIFFERENT TYPES OF CONCRETES BY MONTE CARLO METHODS A. Aminian, Shiraz, Iran M. R. Nematollahi, Shiraz, Iran

SESSION 12C ALTERNATIVE ENERGY (14:00-15:20)

Chairman: Ibrahim Dinçer

PURPOSEFUL SYNTHESIS OF CHEMICAL ELEMENTS AND ECOLOGICALLY PURE MOBILE SOURCES OF ENERGY Vladimir A. Krivitsky, Dubna, Russia Fangil A. Gareev, Dubna, Russia

POTENTIAL OF SOLAR HOME SYSTEMS IN PAKISTAN Mujeebudin Memon, Pakistan Khanji Harijan, Pakistan Mohammad Aslam Uqaili, Pakistan

HEAT AND MASS TRANSFER ANALYSIS IN INTERMEDIATE TEMPERATURE SOLID OXIDE FUEL CELLS (IT-SOFC) Bora Timurkutluk, Niğde, Türkiye Mahmut D. Mat, Niğde, Türkiye Yüksel Kaplan, Niğde, Türkiye

22 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NEW ENERGY OPINION LEADERS' LIFESTYLES AND MEDIA USAGE - APPLYING DATA MINING DECISION TREE ANALYSIS FOR UNIDO-ICHET WEBSITE USERS Mavis Tsai, Taipei, Taiwan Ayfer Veziroglu, Miami, USA Scott Warren, Taipei, Taiwan Yunze Que, Taipei, Taiwan

ENERGY CONUNDRUM, DIGITAL REVOLUTION AND POLITICS Ömer Ersun, Ambassador (Retired), Türkiye

SOCIAL EVENT: BOAT TOUR ON BOSPHORUS WITH GALA DINNER (17:00)

(Departure from World Park Hotel at 16:00)

FRIDAY, 08 JUNE 2007 SESSION 13

SESSION 13A FUSION II (09:00-10:20)

Chairman: Carlos Varandas

D-3HE FUELED FUSION DEVICES AS A STEP TOWARDS TOTAL FUSION SAFETY Laila El-Guebaly, Madison, USA Massimo Zucchetti, Torino, Italy

RECONSTRUCTION AND ANALYSIS OF TEMPERATURE AND DENSITY SPATIAL PROFILES FROM INERTIAL CONFINEMENT FUSION IMPLOSION CORES R. C. Mancini, Reno, USA

STUDY OF CONVERSION EFFICIENCIES FOR LASER FUSION LIKE A VISTA IN THE LASER PLASMA EXPERIMENTS AND PIC-SIMULATIONS Yu. P. Zakharov, Novosibirsk, Russia K. V. Vchivkov, Novosibirsk, Russia A. V. Melekhov, Novosibirsk, Russia V. G. Posukh, Novosibirsk, Russia E. L. Boyarintsev, Novosibirsk, Russia I. F. Shaikhislamov, Novosibirsk, Russia H. Nakashima, Kasuğa-Koen, Japan

ON THE POSSIBILITY OF D-3HE FUSION BASED ON FAST-IGNITION INERTIAL CONFINEMENT SCHEME Yasuyuki Nakao, Fukuoka, Japan K. Hegi, Fukuoka, Japan T. Ohmura, Fukuoka, Japan

23 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

M. Katsube, Fukuoka, Japan T. Johzaki, Osaka, Japan K. Kudo, Fukuoka, Japan M. Ohta, Fukuoka, Japan

SESSION 13B FISSION REACTORS VI (09:00-10:20)

Chairman: Kune Y. Suh

INVESTIGATION OF THE PROPERTIES OF THE NUCLEI USING ON THE NEW GENERATION REACTOR TECHNOLOGY SYSTEMS E. Tel, Ankara, Türkiye H. M. Şahin, Ankara, Türkiye Şenay Yalçın, Istanbul, Türkiye Taner Altınok, Ankara, Türkiye A. Kaplan, İsparta, Türkiye A. Aydın, Kırıkkale, Türkiye

REACTIVITY INSERTION ACCIDENT IN A SMALL MOLTEN SALT REACTOR Yoichiro Shimazu, Japan Nobuhide Suzuki, Japan

COMPUTATIONAL ANALYSIS OF BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM J. Seok Hwang, Seoul, Korea Hyoung M. Son, Seoul, Korea W. Soo Jeong, Seoul, Korea Tae W. Kim, Seoul, Korea Kune Y. Suh, Seoul, Korea

PRELIMINARY STUDY ON CHARACTERISTICS OF EQUILIBRIUM THORIUM FUEL CYCLE OF BWR Abdul Waris, Bandung, Indonesia Rizal Kurniadi, Bandung, Indonesia Zaki Su'ud, Bandung, Indonesia Sidik Permana, Tokyo, Japan

SESSION 13C SOCIETAL ISSUES (09:00-10:20)

Chairman: Masanao Moriwaki

TECHNICAL CONSIDERATIONS ON NUCLEAR TERRORISM Guillemo Velarde, Madrid, Spain

24 13 International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SOCIETY RESPONSE TO NUCLEAR ENERGY Santamaria, Madrid, Spain USEFULLNESS OF UNDER CONTROLLED POWERS O. Alniak, Istanbul, Türkiye

IMPACT OF THE DECOMMISSIONING OF NUCLEAR FACILITIES AND RADIOACTIVE WASTE TRAFFICKING IN AFRICA Baba Gana Abubakar, Maiduguri, Nigeria

COFFEE BREAK (10:20-10:40) SESSION 14

SESSION 14A FUSION III (10:40-12:20)

Chairman: Kazuo Furukawa

INERTIAL CONFINEMENT FUSION AND RELATED TOPICS Alexander N. Starodub, Moscow, Russia

STUDIES ON THE KINETICS OF MUON CATALYZED FUSION IN THE HT MIXTURE WITH VERY LOW CONCENTRATION S. N. Hosseini Motlagh, Tehran -Iran

FRACTAL REACTOR: AN ALTERNATIVE SYSTEM BASED ON NATURE'S GEOMETRY Todd Lael Siler, Colorado, USA

THERMONUCLEAR FUSION BY MECHANICAL ADIABATIC COMPRESSION OF A DENSE PLASMA David W. Kraft, Bridgeport, USA

BENCHMARKING THE CAD-BASED ATTILA DISCRETE ORDINATES CODE WITH EXPERIMENTAL DATA OF FUSION EXPERIMENTS AND TO THE RESULTS OF MCNP CODE IN SIMULATING ITER Mahmoud Z. Youssef, Los Angeles, USA

SESSION 14B NUCLEAR TECHNIQUES (10:40-12:20)

Chairman: Stefan Taczanowski

SUPERCRITICAL CARBON DIOXIDE BRAYTON POWER CONVERSION CYCLE FOR BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM Tae W. Kim, Seoul, Korea

25 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Nam H. Kim, Seoul, Korea Kune Y. Suh, Seoul, Korea

UTILIZING THE SLOWING-DOWN-TIME TECHNIQUE FOR BENCHMARKING NEUTRON THERMALIZATION IN T. Zhou, Raleigh, USA A. I. Hawari, Raleigh, USA B. W. Wehring, Raleigh, USA

SENSITIVITY ANALYSIS OF CORROSION PRODUCT ACTIVITY IN PRIMARY COOLANT OF PWR Muhammad Rafique, Azad Kashmir, Pakistan Naşir M. Mirza, Islamabad, Pakistan Sikander M. Mirza, Islamabad, Pakistan

ANALYSIS AND IMPROVEMENT OF CYCLOTRON THALLIUM TARGET ROOM SHIELD Hajiloo N., Karaj, Iran Raisali G., Tehran, Iran

ENHANCEMENT OF NUCLEAR HEAT TRANSFER IN A TYPICAL PRESSURIZED WATER REACTOR BY NEW SPACER GRIDS Mohammad Nazifi, Shiraz, Iran Mohammadreza Nematollahi, Shiraz, Iran

SESSION 14C (10:40-12:20)

Chairman: Kazuo A. Tanaka

CALIBRATION EXPERIMENTS OF NEUTRON SOURCE IDENTIFICATION AND DETECTION IN SOIL N. V. Gorin, Russia E. N. Lipilina, Russia G. V. Rukavishnikov, Russia D. V. Shmakov, Russia A. I. Ulyanov, Russia

FUSION BY QED CONFINEMENT? T. V. Prevenslik, Berlin, Germany

RELAP5 SCDAP RIA TRANSIENTS ANALYSIS Dr. Gheorghe Negut, Bucharest, Romania

FUSION CHANNEL OF pd CHARGE-SYMMETRIC ION INCLUDING PHOTONS Gheisari, Rouhollah, Boushehr, Iran

LUNCH (12:20-13:30)

26 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SESSION 15

SESSION 15A NUCLEAR FISSION POWER (14:00-15:20)

Chairman: Laila El-Guebaly

PRELIMINARY FEASIBILITY STUDY OF THE HEAT-PIPE ENHS REACTOR Massimiliano Fratoni, Berkeley, USA Lance Kim, Berkeley, USA Sara Mattafirri, Berkeley, USA Robert Petroski, Berkeley, USA Ehud Greenspan, Berkeley, USA

OBSERVATIONS ON THE CANDLE BURN-UP IN VARIOUS GEOMETRIES Walter Seifritz, Hausen, Switzerland

EXPERIMENTAL RESULTS FROM A REACTOR MONITORING EXPERIMENT WITH A CUBIC METER SCALE ANTINEUTRINO DETECTOR Adam Bernstein, Livermore, USA

FEASIBILITY STUDY OF SMALL LONG-LIFE WATER COOLED THORIUM REACTORS (WTRS) FOR PROVIDING SMALL QUANTITY OF ENERGY DEMANDS Ismail, Peng Hong Liem, Tokyo, Japan Sidik Permana, Tokyo, Japan Hiroshi Sekimoto, Tokyo, Japan

SESSION 15B MISCELLANEOUS (14:00-15:20)

Chairman: Andrey V. Gulevich

METAPHYSICS METHODS DEVELOPMENT FOR HIGH TEMPERATURE GAS COOLED REACTOR ANALYSIS Volkan Seker, West Lafayette, USA Thomas J. Downar, West Lafayette, USA

INVESTIGATION OF TRITIUM AND 233U BREEDING IN A FISSION-FUSION HYBRID REACTOR

FUELLING WITH THO2 Kadir Yıldız, Aksaray, Türkiye Necmettin Şahin, Aksaray, Türkiye H. Mehmet Şahin, Ankara, Türkiye Şenay Yalçın, İstanbul, Türkiye Taner Altınok, Ankara, Türkiye Adem Acır, Ankara, Türkiye

27 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Mustafa Bayrak, Niğde, Türkiye Mahmut Alkan, Niğde, Türkiye Okhan Durukan, Niğde, Türkiye

NUMERICAL ANALYSIS OF STEADY STATE FLUID FLOW IN A TWO-DIMENSIONAL WAVY CHANNEL M. Gorji, Babol, Iran E. Hosseinzadeh, Babol, Iran

CRITICAL HEAT FLUX NEAR THE CRITICAL PRESSURE IN HEATER ROD BUNDLE COOLED BY R-134A FLUID: EFFECTS OF UNHEATED RODS AND SPACER GRID Se- Young Chun, Daejeon, Republic of Korea Chan-Whan Shin, Daejeon, Republic of Korea Sung-Deok Hong, Daejeon, Republic of Korea Sang-Ki Moon, Daejeon, Republic of Korea

SESSION 15C LOW ENERGY REACTION AND NUCLEAR PHYSICS (14:00-15:20)

Chairman: Xing Z. Li

INVESTIGATION OF EXCITATION FUNCTIONS USING NEW EVALUATED EMPIRICAL AND SEMI- EMPIRICAL SYSTEMATIC FOR 14-15 MEV (N, T) REACTION CROSS SECTIONS E. Tel, Ankara, Türkiye A. Aydın, Kırıkkale, Türkiye E. G. Aydın, Ankara, Türkiye A. Kaplan, İsparta, Türkiye

LOW ENERGY NUCLEAR REACTIONS: 2007 UPDATE Steven B. Krivit, Los Angeles, USA

EXPERIMENTAL STUDIES ON COLD FUSION AND HYDROGEN METAL Claude Peube-Locou Von Thule, Laroque-Timbaut, France

TUNNELING EFFECT ENHANCED BY LATTICE SCREENING AS MAIN COLD FUSION MECHANISM: A BRIEF THEORETICAL OVERVIEW Fulvio Frisone, Catania, Italy

COFFEE BREAK (15:20-16:00)

CLOSING SESSION (16:00-16:40) & Resume, Final Remarks and Announcement of the 14th ICENES

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(KEYNOTE SPEECH)

OVERVIEW OF PRINCIPLES AND CHALLENGES OF FUSION NUCLEAR TECHNOLOGY

Mohamed Abdou

Distinguished Professor of Engineering and Applied Science And Director of the Center for Energy Science and Technology (CESTAR) University of California-Los Angeles (UCLA), USA E-Mail: [email protected] ABSTRACT Fusion offers very attractive features as a sustainable, broadly available energy source: no emissions of greenhouse gases, no risk of severe accident, and no long-lived radioactive waste. Significant advances in the science and technology of fusion have been realized in the past decades. Seven countries (EU, Japan, USA, Russia, S. Korea, China, and India) comprising about half the world population are constructing a major magnetic fusion facility, called ITER, in France. The objectives of ITER are to demonstrate self-sustaining burning fusion plasma and to test fusion technologies relevant to fusion reactor.

Many challenges to the practical utilization of fusion energy remain ahead. Among these challenges is the successful development of Fusion Nuclear Technology (FNT). FNT includes those fusion system components circumscribing the plasma and responsible for tritium production and processing, heat removal at high temperature and power density, and high heat flux components. FNT components face a new and more challenging environment than experienced by any previous nuclear application. Beyond plasma physics, FNT has most of the remaining feasibility and attractiveness issues in the development of fusion as an energy source. The blanket, a key FNT component, determines the critical path to DEMO. The blanket is exposed to an intense radiation environment. Radioactivity and can be produced in the structure and other blanket elements. Hence, material choices have a large impact on safety and environmental attractiveness. The unique conditions of the fusion environment include high radiation flux, high surface heat flux, strong 3-D-component magnetic field with large gradients, and ultra-low vacuum. These conditions, together with the requirements for high-temperature operation and tritium self-sufficiency, make blanket design and development challenging tasks.

The blanket concepts being considered worldwide can be classified into solid breeders and liquid breeders. Solid breeder concepts use a lithium ceramic as breeder, beryllium as neutron multiplier, and or water as coolant. The liquid breeder can be: a) liquid metal (lithium or S3Pb )7Li), or b) low-conductivity molten lithium salt. Self-cooled, separately-cooled, and dual- coolant concepts are options for liquid breeder blankets. Ferritic steel is considered the primary structural material for almost all blanket concepts for DEMO. There are many advantages and feasibility issues for all blanket concepts. Examples include MHD and insulators, tritium permeation barriers, tritium control, materials interactions and compatibility, thermomechanics interactions, reliability, and synergistic effects. An R&D program is being pursued worldwide that includes testing in non-fusion facilities such as laboratory experiments and fission reactors. Testing of blanket modules in the fusion environment is one of the key objectives of ITER. The

30 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

İTER Parties are presently developing a detailed plan for testing several blanket options on ITER. In addition to ITER, a specialized plasma-based facility for testing and development of FNT is being considered.

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A ROAD MAP FOR THE REALIZATION OF GLOBAL-SCALE THORIUM BREEDING FUEL CYCLE BY SINGLE MOLTEN-FLUORIDE FLOW

Kazuo Furukawa*, Kazuto Arakawa1, L.Berrin Erbay2, Yasuhiko Ito3, Yoshio Kato*, Hanna Kiyavitskaya4, Alfred Lecocq5, Koshi Mitachi6, Ralph Moir7, Hiroo Numata8, J. Paul Pleasant9, Yuzuru Sato10, Yoichiro Shimazu11, Vadim A.Simonenco12, Din Dayal Sood13, Carlos Urban14, Ritsuo Yoshioka15 ^International Thorium Molten-Salt Forum, 1029-5 Oiso, Kanagawa, 255-0003, Japan. 'Res. Cent, for Ultra- High Volt. Electron Microscopy, Osaka Univ., Osaka 567-0047, Japan; 2Mecha. Engi. Dep., Eskişehir Osmangazi Univ., 26480 Bati Meşelik, Eskişehir, Türkiye; 3Dep.of Environ. Systems Sci., Doshisha Univ., Kyoto 610-0321, Japan; 4Joint Inst. for Power & Nucl. Res.-Sosny, Krasin str., 99 220109 Minsk, Belarus; S27ave. des Fauvettes 91400 Orsay, France; 6Dep. of Mecha. Engi., Toyohashi Univ. of Tech., Toyohashi, 441-8580,Japan; 7Vallecitos Molten Salt Res. Associ. 607 E. Vallecitos Rd, Livermore, CA 94550/USA; 8Gradu. Sch. of Metallur. & Ceram. Sci., Tokyo Inst. of Tech., Tokyo, 152-8552,Japan; 9Ene. Frontiers Interna., 11472 Ehren St., Lake View Terrace, CA 91342,USA; 10Dep. of Metallur., Tohoku Univ.,Sendai 980-8579,Japan nDep. of Environ. Systems, Hokkaido Univ., Sapporo 060-0808, Japan; I2Fed. Inst. of Tech. Physics, Snezhinsk, Russia; 13Sector 15, Sanpada, Navi Mumbai 400705, India; 14Thorium Group, Engi. Sch., Fed. Univ. of Minas Gerais and Natio. Nucl. Energy Commission, Brazil; 15Japan Func. Safety Lab., 3-17-24, Hino-chuo, Konan, Yokohama, 234-0053, Japan E-mail: [email protected]

ABSTRACT For global survival in this century, we urgently need to launch a completely new global nuclear fission industry. To get worldwide public acceptance of nuclear energy, improvements are essential not only on safety, radio-waste management and economy but also especially resistance and safeguards. However, such global fission industry cannot replace the present fossil fuel industry in the next 50 years, unless the doubling-time of nuclear energy is less than 10 years, preferably 5-7 years. Such a doubling-time cannot be established by any kind of classical "Fission Breeding Power Station" concept. We need a symbiotic system which couples fission power reactors with a system which can convert fertile thorium to fissile U-233, such as a spallation or D/T fusion (if and when it becomes available).

For such a purpose, THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetic System] has been proposed, which is composed of simple thermal fission power stations (FUJI) and fissile- producing Accelerator Molten-Salt Breeder (AMSB). Its system functions are very ambitious, delicate and complex, but can be realized in the form of simple hardware applying the multi- functional "single-phase molten-fluoride" circulation system. This system has no difficulties relating with "radiation-damage", "heat-removal" and "chemical processing" owing to the simple "idealistic ionic liquid" character. FUJI is size-flexible (economical even in smaller sizes), fuel self-sustaining without any continuous chemical processing and without core-graphite replacement, and AMSB is based on a single-fluid molten-salt target/blanket concept, which solves most engineering difficulties such as radiation-damage, heat-removal etc., except high-current proton accelerator development. Several AMSBs are accommodated in the regional centers (several ten sites in the world) with batch chemical processing plants including radio-waste management. The integrated thorium breeding

32 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye fuel cycle is composed of a simple single-phase molten-fluoride, which is used for all purposes of THORIMS-NES including the transmutation of waste nuclei as a most suitable working medium. This Th-U fuel cycle has significant advantages in negligible production of Trans-uranium elements, nuclear proliferation resistance, economy, etc., and in a high potential for producing hydrogen-fuel by easier high-temperature heat production in future. For its realization we have to develop the following steps successively: (A) MiniFUJI (7-10MWe): laying foundation for the basic MSR technology and specialists, reconfirming/improving the ORNL-MSR-Program results, the successful 4 years operation experience of MSRE (Molten-Salt Reactor Experiment); (B) FUJI-Pu (100-300MWe): incinerating initial containing MS-fuel prepared easily by dry processing (simplified FREGATE project without solid-fuel reproduction) from spent solid- fuels of existing nuclear power stations, and producing U233. It means smooth/gradual shifting from present U-Pu cycle era; (C) AMSB: producing U233 depending on the matured MSR technology 20-30 years later. (D) THORIMS-NES: globally deploying the regional centers.

Now the real Th-U Breeding Fuel-cycle would be implemented for global survival in the issues of not only energy, environment and poverty but also the perfect elimination of the wars through the extinction of nuclear weapons by nuclear burnup of materials, for which purpose the Th-U cycle has a significant advantage over the U-Pu cycle. In this study all items mentioned are explained, evaluated and discussed thoroughly for the sake of global survival.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

THE NATIONAL IGNITION FACILITY (NIF): A PATH TO FUSION ENERGY

Edward I. Moses Lawrence Livermore National Laboratory, Livermore, CA 94550, E-Mail: [email protected]

ABSTRACT Fusion energy has long been considered a promising clean, nearly inexhaustible source of energy. Power production by fusion micro-explosions of inertial confinement fusion (ICF) targets has been a long term research goal since the invention of the first laser in 1960. The NIF is poised to take the next important step in the journey by beginning experiments researching ICF ignition. Ignition on NIF will be the culmination of over thirty years of ICF research on high-powered laser systems such as the laser at LLNL and the OMEGA laser at the University of Rochester as well as smaller systems around the world. NIF is a 192 beam Ndglass laser facility at LLNL that is more than 90% complete. The first cluster of 48 beams is operational in the laser bay, the second cluster is now being commissioned, and the beam path to the target chamber is being installed. The Project will be completed in 2009 and ignition experiments will start in 2010. When completed NIF will produce up to 1.8 MJ of 0.35 \xm light in highly shaped pulses required for ignition. It will have beam stability and control to higher precision than any other laser fusion facility. Experiments using one of the beams of NIF have demonstrated that NIF can meet its beam performance goals. The National Ignition Campaign (NIC) has been established to manage the ignition effort on NIF. NIC has all of the research and development required to execute the ignition plan and to develop NIF into a fully operational facility. NIF will explore the ignition space, including direct drive, 2co ignition, and fast ignition, to optimize target efficiency for developing fusion as an energy source. In addition to efficient target performance, fusion energy requires significant advances in high repetition rate lasers and fusion reactor technology. The Mercury laser at LLNL is a high repetition rate Nd-glass laser for fusion energy driver development. Mercury uses state-o-the art technology such as ceramic laser slabs and light diode pumping for improved efficiency and thermal management. Progress in NIF, NIC, Mercury, and the path forward for fusion energy will be presented.

* This work was performed under the auspices of the U. S. Department of Energy by the University of California Lawrence Livermore National Laboratory under contract No. W-7405- Eng-48.

34 SESSION 2: PLENARY SESSION TR0700274

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SPACE NUCLEAR REACTOR CONCEPTS FOR AVOIDANCE OF A SINGLE POINT FAILURE

Mohamed S. El-Genk Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept. The University of New Mexico, Albuquerque, NM 8713, USA, (505) 277 - 5442, E-Mail: [email protected]

ABSTRACT This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors.

36 TR0700275

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

IRRADIATION FACILITIES FOR MATERIALS RESEARCH: IFMIF AND SMALL SCALE INSTALLATIONS

Jose M Perlado, M. Victoria Institute» Fusion Nuclear (DENIM) / ETSII / Universidad Politecnica de Madrid J. Gutierrez Abascal, 2; 28006 Madrid E-mail: [email protected]

ABSTRACT The research of advance materials in nuclear fields such as new fission reactors (Generation-IV), Accelerator Driven Systems for Transmutation of Radioactive Wastes and Nuclear Fusion, is becoming very much common in the types of low activation and radiation resistant Materials. Ferritic-Martensitic Steels (based in 9-12 Cr) with or without Oxide Dispersion Techniques (Ytria Nanoparticles), Composites materials are becoming the new generation to answer requirements of high temperature, high radiation resistance of structural materials. Special dedication is appearing in general research programmes to this area of Materials. The understanding of their final performance needs a wider knowledge of the mechanisms of radiation damage in these materials from the atomistic scale to the macroscopic responses. New extensive campaigns are being funded to irradiate from simple elements to model alloys and finally the complex materials themselves. That sequence and its state of art will be presented

One clear technique for that understanding is the Multiscale Modelling which includes simulation techniques from quantum mechanics, molecular dynamics, defects diffusion, mesoscopic modelling and finally the macroscopic constitutive relations for macrosciopic analysis. However, in each one of these steps is necessary a systematic and well established program of experiments that combines the irradiation and the very detailed analysis with techniques such as Transmission Electron Microscope, Positron Annihilation, SIMS, Atom Probe, Nanoindebntation

A key aspect that wants to be presented in this work is the state of art and discussion of Irradiation Facilities for Materials studies. Those facilities goes from ion implantation sources, small accelerator, Experimental Reactors such High Flux Reactor, sophisticated Triple Beams Sources as JANNUS in France to generate at the same time displacements-hydrogen-helium, and projected very large neutron installation such as IFMIF. The role to be covered for each one will be presented. Particular attention will be dedicated in this work to the D-Li stripping reaction very Intense Source of Neutron Irradiation, International Fusion Materials Irradiation Facility (IFMIF), being designed by Europe and Japan under funding from the Broader Approach, and where Spain will play a major role inside Europe. Its design will be reviewed together with the major questions and the comparison with other potential ideas for Neutron Sources for Engineering development of Materials submitted to irradiation.

37 TR0700276 !I . _„ . _ i.. 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SUSTAINABLE ENERGY SYSTEMS AND THE EURATOM RESEARCH PROGRAMME

S. Webster and G. Van Goethem European Commission, Research Directorate-General, Directorate J - Energy (Euratom), Brussels, Belgium E-Mail: simon. webster@cec. eu. int; georges. van-goethem@cec. eu. int

ABSTRACT We are at a turning point in European research. With the launch of the EU's 7th Framework Programme, committing some €53 billion of public funds to the European research effort over the next 7 years, Europe has finally woken up to the importance of R&D in the realisation of the most fundamental objectives defining the Union: growth, competitiveness, and knowledge. At the same time, and with strong links to growth and competitiveness but also to environmental protection, the Union is in the throws of an intense debate on future energy policy and climate change. Part of the research budget, some would say too small a part, is earmarked for energy - in particular the technological aspects of low carbon systems such renewables. This effort, together with measures to improve the EU's security and independence of supply, are essential if Europe is to respond effectively to solve the future energy conundrum. But where does nuclear fit in all this? What will the Union be doing in the area of nuclear research? Indeed, does nuclear figure at all in the long-term plans of the Union? Through the Euratom part of the Framework Programme, the EU is maintaining important support to up-stream research in the area of advanced reactor technologies. This effort is being coordinated at the global level through Euratom's membership of the Generation-IV International Forum. Though EU research in this field still has its critics among the Member States, and despite the relatively small sums currently committed, the leverage effect of current actions is significant and this is set to grow in the future. The imminent setting up of a Strategic Energy Technology Plan, as part of the European Commission on-going activities in the field of energy policy, and the feedback from independent experts in the Advisory Group on Energy and the Euratom Scientific and Technical Committee all point to following conclusions: EU support for research on advanced nuclear fission technology is justified as part of a broad portfolio approach to R&D on new energy systems and carriers, this technology is extremely promising as part of the answer to our long-term energy problems, and the current level of R&D support is insufficient in view of the challenges faced.

38 TR0700277

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

R&D ACTIVITIES OF THE JOINT RESEARCH CENTRE (JRC) AND ITS INVOLVEMENT IN THE DEVELOPMENT OF FUTURE NUCLEAR ENERGY SYSTEMS

Roland Schenkel (Director General of the JRC), Germay E-Mail: roland. schenkel@ec. europa. eu

ABSTRACT Besides the policy driven support which the JRC gives to the European Commission and its Member States, the nuclear activities of the JRC also fulfil the R&D obligations as enshrined in the Euratom Treaty. These have for objectives to develop and assemble knowledge in the field of nuclear energy and concern basic actinide research, nuclear data and nuclear measurements, radiation monitoring and radionuclides in the environment, health and , management of spent fuel and waste, safety of reactors and fuel cycle and nuclear safeguards and non proliferation. The European Union currently imports 50% of its energy and, going by the present trend, this may increase to 70% within 20 years. One third of the electricity in Europe is currently been produced via nuclear fission and the move to innovative reactor systems holds great promise. In May 2006, the European Atomic Energy Community became a Party to the Framework Agreement for International Collaboration on Research and Development of Generation IV Nuclear Energy Systems (GIF Framework Agreement). The "Generation IV" initiative concerns concepts for nuclear energy systems that can be operated in a manner that will provide a competitive and reliable supply of energy, while satisfactorily addressing nuclear safety, waste, proliferation and public perception concerns. The JRC with its strong international dimension is not only the implementing agent for Euratom in the Generation IV international forum, but also participates actively in related R&D projects. The R&D projects are focused on fuel development, reprocessing and irradiation testing, fuel- cladding interaction and corrosion, basic data for fuel and reprocessing, reprocessing and waste treatment. In this paper the R&D the nuclear activities of the JRC will be presented especially those related to its participation to GIF.

39

PARALLEL SESSION 3A: THE FUTURE OF NUCLEAR FISSION TR0700278

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

INNOVATIVE FISSION REACTORS FOR THIS CENTURY

Emilio MINGUEZ Professor Chair in Nuclear Technology Universidad Politecnica de Madrid (UPM), Spain E-Mail: [email protected]

ABSTRACT It is well known that global trends indicate a rebirth of nuclear energy due to several items: the climate change and the use of energies that emits CO2, the cost and dependence of gas and oil, the new innovative reactors which are competitive, safer, and sustainable and can support the Kyoto Protocol. The Advanced Reactors have safer systems than those developed in the Generation II, which demonstrates that are sustainable for the present and nuclear industry has also developed new concepts for the future which also will be sustainable. Now the new power plants that have being constructed are classified in the Generation III. Several units of this technology are in operation in Japan and other countries of the Pacific. Europe is now constructing the first unit in Finland (Olkilouto) with European technology: the European Pressurized Reactor (EPR). France has announced the beginning of the construction of an EPR in Flamanville next year. In 2000, several countries with advanced nuclear technology established the Generation IV International Forum (GIF) to develop and demonstrate nuclear energy systems that offer advantages in the following areas: sustainability, economics, safety and reliability and proliferation resistance and physical protection. These new systems will be deployed commercially after 2030. Six innovative concepts are under research, and the aim is not only produce electricity, but also hydrogen using the operational conditions of several concepts. Developed countries with NPPs in operation have strategies for the future of the nuclear energy. For the short term is to extend the operation of the NPPs until 60 years, or alternatively construction of new units of Generation III, to substitute those closed for decommissioning, keeping the percentage of contribution to the electricity generated. Between the period 2030-50, the solution is to operate the new innovative systems of the Generation IV, which uses the passive concept, and in the second part of the 21st Century both innovative fission reactors and fusion reactors. For 2025, it seems that many countries of EU will have to construct NPPs until 40 GWe: France, UK, Germany, North Europe, Russia, Spain, Rumania and Türkiye, between others. The viability of these innovative concepts will be presented in this paper.

42 TR0700279

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NUCLEAR POWER IN SPACE

Samim Anghaie Professor & Director Innovative Nuclear Space Power & Propulsion Institute University of Florida 202 NSC, Gainesville, FL 32611-8300, USA Tel: (352) 392-8653, Fax,: (352) 392-8656 E-mail: [email protected]

ABSTRACT The development of space nuclear power and propulsion in the started in 1955 with the initiation of the ROVER project. The first step in the ROVER program was the KIWI project that included the development and testing of 8 non-flyable ultrahigh temperature nuclear test reactors during 1955-1964. The KIWI project was precursor to the PHOEBUS carbon-based fuel reactor project that resulted in ground testing of three high power reactors during 1965-1968 with the last reactor operated at 4,100 MW. During the same time period a parallel program was pursued to develop a nuclear based on cermet fuel technology. The third component of the ROVER program was the Nuclear Engine for Rocket Vehicle Applications (NERVA) that was initiated in 1961 with the primary goal of designing the first generation of nuclear based on the KIWI project experience. The fourth component of the ROVER program was the Reactor In-Flight Test (RIFT) project that was intended to design, fabricate, and flight test a NERVA powered upper stage engine for the Saturn-class lunch vehicle. During the ROVER program era, the Unites States ventured in a comprehensive space nuclear program that included design and testing of several compact reactors and space suitable power conversion systems, and the development of a few light weight heat rejection systems. Contrary to its sister ROVER program, the space nuclear power program resulted in the first ever deployment and in-space operation of the nuclear powered SNAP-10A in 1965.

The USSR space nuclear program started in early 70's and resulted in deployment of two 6 kWe TOPAZ reactors into space and ground testing of the prototype of a relatively small nuclear rocket engine in 1984. The US ambition for the development and deployment of space nuclear powered systems was resurrected in mid 1980's and intermittently continued to date with the initiation of several research programs that included the SP-100, Space Exploration Initiative, the Jupiter Icy Moons Orbiter (JIMO), the project PROMETHUS, and the Moon and surface power project.

A brief history, future prospect, and some technology challenges of space nuclear power and propulsion will be discussed.

43 TR0700280

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

THE FIXED BED NUCLEAR REACTOR CONCEPT

Sümer Şahin Gazi University, Türkiye E-Mail.'sumer@gazi. edu. tr

Farhang Sefidvash Federal University of Rio Grande do Sul, Porto Alegre, Brazil E-y[&il:farhang@ufrgs. br

ABSTRACT The core of a water moderated Fixed Bed Nuclear Reactor (FBNR), possessing, for instance, an electrical power of 40 MW, consists of 1.35 million fuel pellets (9.5t) with a diameter of 1.5 cm each. The low fuel is made of TRISO type microspheres used in the HTGR, embedded in a graphite matrix and cladded by a shell of lmm SiC.

Under any thinkable operational condition the fuel temperature will be below 400 C whereas ist stability limit is at about 1600 C. The first characteristic of the FBNR is, therefore, its robust fuel under relatively "cold" operating conditions and - due to the outer SiC - shell layer - the freedom from any hydrogen production.

To operate the reactor the fuel pellets are pumped by a flow of water from below into the core regions where they form a stable fixed bed of about 4 cubic meter and become critical for energy production heating the outlet water to about 330 C (at 160 bar) which feeds a steam generator.

The new safety feature is now the following: In case of any abnormity (e.g. external power failure, overheating etc.) the circulating pump stops and - due to gravity - the fuel pellets fall automatically out of the core region into a helical "fuel chamber" underneath the core where their decay heat is transferred passively by natural circulation to a water tank housing the fuel chamber. The safety principle, applied here, is: The loss of an active component (circulating pump) induces a self-controlled, passively working shut-down maneuvre accompanied by a foolproof decay heat removal without any emergency power system or any human interaction.

The fuel chamber is sealed and is transported as the only reactor component to and from the reactor site. There is no possibility to irradiate fertile fuel, too. For a long-life core (larger than a 10 years cycle time) the fuel can either be poisoned by gadolinium-oxide or by a piston type core limiter adjusting the height and controlling thereby the number of the fuel pellets in the active core region. Refueling is done only by changing the fuel chamber allowing also the use of this foolproof light water reactor in countries without a perfect nuclear infrastructure.

In a conclusion we will present a completely new pressurized water reactor, based on a robust fuel and a purely passively working decay heat removal system. Aside from electricity production in the realm of 10-60 MW (electrical) it can also be used for water desalination and space heating purposes.

The criticality calculations are made using SCALE codes. The SCALE system has the capacity to perform criticality, shielding, radiation source term, spent fuel depletion/decay, and heat

44 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye transfer analysis using well established functional modules tailored to the SCALE system. The CSAS control module contains criticality safety analysis sequences that calculate the neutron multiplication factor a multidimensional KENO V.a system models.

45 TR0700281

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

CHALLENGES OF NUCLEAR POWER FOR THE SUSTAINABLE ROLE IN KOREAN ENERGY POLICY

Young Eal LEE 103-16, Munji-dong, Yusung-gu, Daejeon, 305-380, Korea Korea Electric Power Research Institute E-mail: yelee@kepri. re. fcr.Tel: 82-42-865-5174

ABSTRACT This study aims to introduce the current role of nuclear power of Korea as the economic and low carbon emitter in the long term expansion planning and to improve the public acceptance of nuclear as the environmentally friendliness energy source. Nuclear and coal have been selected as the major electricity sources due to the insufficient domestic energy resources, and will provide more than 60% of total electricity generation in Korea for quite some time. National energy policy addressing environmental friendliness, stable supply and least cost has made it difficult to decide which energy resource is the best for the long term energy planning. Climate change regime will diminish the coal power plants in generation amount, the public still keeps nuclear at a distance and insists to replace nuclear by renewable and renewable doesn't any guarantee of stable supply although its economics is fast being improved. Therefore, it is necessary to analyze the long-term power expansion planning in various points of view such as environmental friendliness, benefit of carbon reduction and system reliability as well as least cost operation. The objective and approach of this study are to analyze the proper role of nuclear power by comparing the different types of scenarios in terms of the system cost changes, CO2 emission reduction and system reliability. The results from this analysis are useful for the Korean government in charge of long-term energy policy to go over what kinds of role can each electric resources play and what are the best way to solve the triangular dilemma as economics, environmental friendliness, stable supply of the electricity.

46 TR0700282

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

REACTOR PHYSICS IDEAS TO DESIGN NOVEL REACTORS WITH FASTER FISSILE GROWTH

V. Jagannathan, Usha Pal, R. Karthikeyan, Devesh Raj, Argala Srivastava and Suhail Ahmad Khan* Light Water Reactors Physics Section, Reactor Physics Design Division, *Reactor Project Division Bhabha Atomic Research Centre, Mumbai - 400 085, India E-mail: vjaganl952@rediffinail. com

ABSTRACT There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 235U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities.

47

PARALLEL SESSION 3B: ACCELERATOR DRIVEN SYSTEMS I TR0700283 |

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

DEVELOPMENT OF THE HEAT SINK STRUCTURE OF A BEAM DUMP FOR THE PROTON ACCELERATOR

Wan Young Maeng, C. S. Gil, J. H. Kim, D. H. Kim Korea Atomic Energy Research Institute E-Mail: [email protected]

ABSTRACT The beam dump is the essential component for the good beam quality and the reliable performance of the proton accelerator. The beam dump for a 20 MeV and 20 mA proton accelerator was designed and manufactured in this study. The high heats deposited, and the large amount of radioactivity produced in beam dump should be reduced by the proper heat sink structure. The heat source by the proton beam of 20 MeV and 20 mA was calculated. The radioactivity assessments of the beam dump were carried out for the economic shielding design with safety. The radioactivity by the protons and secondary in designed beam dump were calculated in this sturdy.

The effective engineering design for the beam dump cooling was performed, considering the mitigation methods of the deposited heats with small angle, the power densities with the stopping ranges in the materials and the heat distributions in the beam dump. The heat sink structure of the beam dump was designed to meet the accelerator characteristics by placing two plates of 30 cm by 60 cm at an angle of 12°. The highest temperatures of the graphite, copper, and copper faced by cooling water were designed to be 223°, 146°, and 85°, respectively when the velocity of cooling water was 3 m/s.

The heat sink structure was manufactured by the brazing graphite tiles to a copper plate with the filler alloy of Ti-Cu-Ag. The brazing procedure was developed. The tensile stress of the graphite was less than 75% of a maximum tensile stress during the accelerator operation based on the analysis. The safety analyses for the commissioning of the accelerator operation were also performed. The specimens from the brazed parts of beam dump structure were made to identify manufacturing problems. The soundness of the heat sink structure of the beam dump was confirmed by the fatigue tests of the brazed specimens of the graphite-copper tile components with the repetitive heating and cooling.

The heat sink structure developed in this study can be applied to plasma facing components, collimators of accelerator and dumps for experimental targets to resolve high heat loads and to reduce high radioactivity.

50 TR0700284

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SUBCRITICALITY DETERMINATION IN ADS: THE YALINA-BOOSTER EXPERIMENTS

Carl-Magnus Persson1, Waclaw Gudowski Dept. of Nuclear and Reactor Physics, Royal Institute of Technology (KTH) S-106 91 Stockholm, Sweden E-mail: [email protected]

Andrei Fokau2, Victor Bournos, Yurii Fokov, Christina Routkovskaia, Ivan Serafimovich, Hanna Kiyavitskaya Joint Institute for Power and Nuclear Research - Sosny National Academy of Sciences of Belarus 220109 Minsk, Belarus E-mail: [email protected]

ABSTRACT Introduction A major problem in operating a full-scale subcritical accelerator-driven system (ADS) is to ensure sufficient margin to criticality. Therefore, reliable techniques for subcriticality monitoring are required. In order to develop such techniques, a full understanding of existing reactivity determination methods is essential. In this work, reactivity determination methods, such as pulsed neutron source methods and noise methods, are studied experimentally in the subcritical facility YALINA-Booster.

YALINA-Booster The subcritical assembly YALINA-Booster recently constructed at the Joint Institute for Power and Nuclear Research - Sosny, consists of a subcritical core driven by an external neutron source. The neutron source is a powerful consisting of a deuteron accelerator and a target of or tritium embedded in titanium. Through (d, d) - or (d, t)-reactions neutrons are created with energy around 2.5 MeV and 14.1 MeV respectively. Neutrons are born in the centre of the core and multiply through a lead matrix fuelled with highly enriched uranium (90% and 36%). This zone is referred to as the booster zone and is surrounded by a thermal zone, moderated by polyethylene. In order to reach sufficient high effective multiplication factor, the thermal zone is fuelled by approximately one thousand rods of 10% enriched in cylindrical geometry. To prevent thermal neutrons from diffusing into the fast booster zone, an interface, consisting of carbide and natural uranium rods, is located between the zones. YALINA-Booster has a radial graphite reflector of thickness 24 cm (Figure 1).

Experiments Experiments using the neutron source in pulsed mode will be presented, relying on methods such as the area method and the method of prompt neutron decay rate determination. Moreover, results from noise analysis using for instance the Feynman-a method will be presented.

51 TR0700285

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

STUDY OF FISSION CROSS SECTIONS INDUCED BY NUCLEONS AND PIONS USING THE CASCADE-EXCITON MODEL CEM95

Zafar Yasina'\ M. ikram Shahzad" aDepartment of Nuclear Engineering, PIEAS, P.O. Nilore, Islamabad, Pakistan bPhysics Research Division, PINSTECH, P.O. Nilore, Islamabad, Pakistan E-Mail: [email protected]; [email protected]

ABSTRACT Nucleon and pion-induced fission cross sections at intermediate and at higher energies are important in current nuclear applications, such as accelerator driven-systems (ADS), in medicine, for effects on electronics etc. In the present work, microscopic fission cross sections induced by nucleons and pions are calculated using the cascade-exciton model code CEM95 for different projectile-target combinations; at various energies and the computed cross sections are compared with the experimental data found in literature. A new approach is used to compute the fission cross sections in which a change of the ratio of the level density parameter in fission to neutron emission channels was taken into account with the change in the incident energy of the projectile. We are unable to describe well the fission cross sections without using this new approach. Proton induced fission cross sections are calculated for targets !97Au, 208Pb, 209Bi, 238U and 239Pu in the energy range from 20 MeV to 2000 MeV. Neutron induced fission cross sections are computed for 238U and 239Pu in the energy range from 20 MeV to 200 MeV. Negative pion induced cross sections for fission are calculated for targets 197Au and 208Pb from 50 MeV to 2500 MeV energy range. The calculated cross sections are essential to build a data library file for accelerator driven- systems just like was built for conventional nuclear reactors. The computed values exhibited reasonable agreement with the experimental values found in the literature across a wide range of beam energies. Telephone: +92-51-2207381 FAX: +92-51-9223727

52 TR0700286

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

MONTE CARLO STUDIES IN ACCELERATOR-DRIVEN SYSTEMS FOR TRANSMUTATION OF HIGH-LEVEL NUCLEAR WASTE

Başar Şarer1, M. E. Korkmaz1, M. Günay2, A. Aydin3 !Gazi Üniv. Fen-Edeb. Fak. Fizik Böl. ANKARA 2İnönü Üniv. Fen-Edeb.Fak. Fizik Böl. MALATYA 3Kırıkkale Üniv. Fen-Edeb. Fak. Fizik Böl. ANKARA E-mail: sarer(a),zazi.edu.tr

ABSTRACT A spallation neutron source was modeled using a high energy proton accelerator. The aim of the core design is to optimise the core parameters for maximizing the minor actinides and fission products transmutation rates, which is created from the operation of nuclear power reactors for the production of electricity, while maintaining the structural material damage and decay heat as low as possible.

The transmutation system is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1 GeV proton beam incident on a natural lead cylindrical target, 20 cm radius, 70 cm height, and entering the target through a 5.3 cm radius hole. The protons were uniformly distributed across the beam of radius 2 cm. The core is cylindrical assembly, 2.3 m radius, 4.6 m high. The wall thickness of the main vessel is 2 cm. The main vessel surrounded by a reflector made of graphite, 40 cm thick. The axes of proton beam and target are concentric with the main vessel axis. The structural walls and beam window are made of the same material, stainless steel, HT9. All dimensions of systems are results of target and core optimization that keeps most of the spallation neutrons within the lead target and transmutes the largest fraction of the long-lived waste. We investigated the following neutronics parameters with presence and absence of fissile materials: • spallation neutron and other particles such as proton, pions and müons yields (per one incident proton) from the spallation target, • spatial and energy distribution of the spallation neutrons, and protons in target, • heat deposition distribution in the spallation target, • heat deposition in beam window, core, reflector and structural components, • transmutation rate of minor actinides and fission products, • atomic displacement and the production rates of hydrogen, helium, deuterium and tritium in the natural lead target, beam window, core and structural components, In the calculations, the Monte Carlo code MCNPX, which is a combination of LAHET and MCNP, was used. The neutron interactions were calculated by the MCNP code below 20 MeV. To transport a wide variety of particles, The Los Alamos High Energy Transport Code (LAHET) was used. These include charged particles such as protons and charged pions as well as neutrons with high energies. MCNPX offers options based on three physics packages; the BERTINI and ISABEL models taken from the LAHET Code System, and the CEM package.

53 TR0700287 |

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

CALCULATIONS of THE MAIN FREE PATH on NEUTRON EMISSION CROSS- SECTION for SPALLATION REACTION of TARGET and FUEL NUCLEI

Eyyüp Tel1, H. F. Kışoğlu2, A. Kayış Topaksu2, A. Aydın3, A. Kaplan4 1 Gazi University, Faculty of Arts and Science, Department of Physics, Beşevler, Ankara-Türkiye 2 Çukurova University, Faculty of Arts and Science, Department of Physics, Yüreğir, Adana- Türkiye 3Kırıkkale University, Faculty of Art and Science, Kırıkkale, Türkiye 4 Süleyman Demirel University, Science and Letter Faculty, Department of Physics, Isparta- Türkiye E-mail: [email protected]

ABSTRACT There are several new technological application fields of fast neutrons such as accelerator-driven incineration/ transmutation of the long-lived radioactive nuclear wastes (in particular transuranium nuclides) to short-lived or stable isotopes by secondary spallation neutrons produced by high-intensity, intermediate-energy, charged-particle beams, prolonged planstary space missions, shielding for particle accelerators. Especially, accelerator driven subcritical systems (ADS) can be used for fission energy production and /or nuclear waste transmutation as well as in the intermediate-energy accelerator driven neutron sources, ions and neutrons with energies beyond 20 MeV, the upper limit of exiting data files that produced for fusion and fission applications. In these systems, the neutron scattering cross sections and emission differential data are very important for reactor neutronics calculations. The transition rate calculation involves the introduction of the parameter of mean free path determines the mean free path of the nucleon in the nuclear matter. This parameter allows an increase in mean free path, with simulation of effect, which is not considered in the calculations, such as conservation of parity and angular momentum in intra nuclear transitions. In this study, we have investigated the multiple pre- equilibrium matrix element constant from internal transition for Uranium, Thorium, (n,xn) neutron emission spectra. The neutron-emission spectra produced by {n,xn) reactions on nuclei of some target (for spallation) have been calculated. In the calculations, we have used the geometry dependent hybrid model and the cascade exciton model including the effects of the pre- equilibrium. The pre-equilibrium direct effects have been examined by using full exciton model. All calculated results have been compared with the experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other. Keywords: Cross-section, pre-equilibrium models, exciton model, main free path and emission spectra

54 PARALLEL SESSION 3C: ALTERNATIVE ENERGY TR0700288

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

ARE THE LESSONS OF THE TECHNO-ECONOMIC STUDIES ON THE SULPHUR- IODINE CYCLE APPLICABLE TO THE OTHER CYCLES?

F. Werkoff,C.Mansilla CEA/SACLAY-DEN/DM2S/SERMA Bât 470 -91191 GIF-SUR-Yvette CEDEX- FRANCE E-mail: francois. [email protected]

ABSTRACT Further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. To explore these opportunities, several countries are involved in a forum on the development of next generation nuclear energy systems known as "Generation IV". Six concepts have been chosen by the forum, to be studied. The Very High Temperature Reactor (VHTR) offers perspectives for producing electricity and hydrogen with high efficiencies. Nuclear production of hydrogen by thermochemical means is one of the prime candidates for powering the hydrogen economy without producing green house gases. Among them, the Sulphur-Iodine (S-I) thermochemical cycle appeared well fitted with the VHTR, due to the temperature needed for the decomposition of the sulphuric acid. It was invented in the 1970's and it benefits from a revival of interest in the framework of Generation IV.

In the last past years, assessments of the S-I process, coupled with a VHTR have been carried out. It appeared that these assessments have to be considered, looking with a particular care to the recommendations of the Generation IV crosscutting economics group [1]: a Generation IV system will:

1. Have a clear life-cost advantage over other energy systems. 2. Have a level of financial risk comparable to other energy projects.

The experience gained from techno-economic studies [2, 3] which consider the S-I cycle, indicates that the choice of alternatives cycles to the S-I one must be driven by ^he characteristic of a previously selected nuclear reactor, mainly the temperature at the nuclear core outlet. Moreover, the net efficiency of the thermochemical cycle must be higher than a reference value defined from the alkaline electrolysis fed by the electricity produced from the selected reactor.

Besides, the technical feasibility of the thermochemical processes is not yet established and the production cost of hydrogen from these processes is the result of the sum of several cost factors which are not sufficiently known at present, but can be expressed through the use of fuzzy sets. The main costs factors would be: investments, raw elements, energy and maintenance costs. The competitiveness of a thermochemical cycle by around 2030 needs to be assessed by estimating the production cost of 1 kg of hydrogen, taking into account the uncertainty linked to the cost factors.

References [1] Generation IV Technology Roadmap (http://gif.inel.gov/roadmap/).

56 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

[2] F.Werkoff, On the profitability of hydrogen of production using nuclear power, CD- Proceedings of the 15* World Hydrogen Energy Conference, Yokohama, Japan, June 27-July 2, 2004, paper 1091. [3] T.Baerecke et al. On the use of genetic algorithms and fuzzy logic for the assessment of the cost of hydrogen to be produced by advancedprocesses. CD-Proceedings of the 2006 Fuel Cell Seminar Hydrogen Energy Conference, Honolulu, November 13-17, 2006, paper 223.

57 TR0700289

13* International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

PRODUCTION METHOD OF HYDROGEN BY JET PLASMA PROCESS IN HYDRO MACHINERY

Farzan Amini

Department of Mechanical Engineering, Farab Company, No.30, Mirhadi St. Vali-Asr Ave.Tehran - Iran E-mail: [email protected] ABSTRACT The purpose of present paper is to the process of plasma formation in hydro machinery when a hydro turbine operates at various conditions and load rejection. By investigation the power, shock pressure , and impact effects of hydro machinery, it is revealed that energy and hydrogen are generated by the plasma process. The investigation on several turbines of various hydro power plants reveals that cold fusion process in hydro machinery generates hydrogen. The hypothesis concerning the participation of alkaline metals in river water and the atomic nuclei of the runner blade material in the formation of hydrogen are considered. It is possible to assume hydrogen, deuterium, helium, and tritium atoms (based on Dr. Mizuno and Dr. Kanarev theories) that are formed, diffuse into cavitation bubbles. The plasma is generated during the collapse of the bubble; thus, the quantity of burnt hydrogen determine the volume of generating hydrogen and the impact force caused by hydrogen explosion (noise).There are five main notions, which can determine hydrogen and plasma process: (1) turbine power effect, (2) high shock pressure, (3) crack on turbine parts, (4) impacts effects and (4) the lift of rotating parts. The frequency of the excitation lies in a range from 0.786 to 1.095 Hz.In future, it may be possible to design hydro turbines based on the plasma process that generates hydrogen; or there may exist turbines that rotate with a mixture of hydrogen explosion and water energies. %

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58 TR0700290

13* International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

ANALYSIS OF AN INNOVATIVE SOLAR WATER DESALINATION SYSTEM USING GRAVITY INDUCED VACUUM.

Teoman AYHAN*, Husain Al-Madani Department of mechanical Engineering, College of Engineering, University of Bahrain, Bahrain Fax: 973 17684844, E-mail: [email protected] Fax: 973 17684844, E-mail: [email protected]

ABSTRACT This study presents the theoretical analysis, design and appropriate models of a new desalination system using gravity induced vacuum. The system utilizes natural means (gravity and atmospheric pressure) to create a vacuum under which water can be rapidly evaporated at much lower temperatures with less energy than conventional techniques. This technique is developed to overcome water storage, in the areas where good solar radiation (or waste heat sources) and sea water (or waste water sources). The developed system consists of an evaporator connected to condenser by means of a vacuum tank. The vapour produced in the evaporator is driven to condenser through the vacuum tank, where it condenses and collected as a product. Vacuum equivalent to 7 kPa (abs) or less can be created depending on ambient temperature of Bahrain climatic conditions. The effect of various operating conditions, namely water levels in condensation and evaporating columns on the system performance were studied. The theoretical analysis and preliminary experimental results show that the performance of this system depends on the condensation temperature.

59 TR0700291 i

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

EXPERIMENTAL INVESTIGATION TWO PHASE FLOW IN DIRECT METHANOL FUEL CELLS

Mahmut D. Mat, Yüksel Kaplan, Selahattin Çelik, Aytekin Öztorul Niğde Üniversitesi, Mühendislik Mimarlik Fak. Makine Müh. Bölümü, 51100 Niğde E-mail: mdmat@Niğde.edu.tr

ABSTRACT Direct methanol fuel cells (DMFC) have received many attentions specifically for portable electronic applications since it utilize methanol which is in liquid form in atmospheric condition and high of the methanol. Thus it eliminates the storage problem of hydrogen. It also eliminates humidification requirement of polymeric membrane which is a problem in PEM fuel cells. Some electronic companies introduced DMFC prototypes for portable electronic applications. Presence of carbon dioxide gases due to electrochemical reactions in anode makes the problem a two phase problem. A two phase flow may occur at cathode specifically at high current densities due to the excess water. Presence of gas phase in anode region and liquid phase in cathode region prevents diffusion of fuel and oxygen to the reaction sites thus reduces the performance of the system. Uncontrolled pressure buildup in anode region increases methanol crossover through membrane and adversely effect the performance. Two phase flow in both anode and cathode region is very effective in the performance of DMYC system and a detailed understanding of two phase flow for high performance DMFC systems. Although there are many theoretical and experimental studies available on the DMFC systems in the literature, only few studies consider problem as a two-phase flow problem.

In this study, an experimental set up is developed and species distributions on system are measured with a gas chromatograph. System performance characteristics (V-I curves) is measured depending on the process parameters (temperature, fuel ad oxidant flow rates, methanol concentration etc).

60 TR0700292

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

POTENTIAL OF HYDROGEN PRODUCTION FROM WIND ENERGY IN PAKISTAN

Mohammad Aslam Uqailia'*, Khanji Harijanb and Mujeebuddin Memonc a Professor, Department of Electrical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan PhD Student, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan c Professor, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan E-mail: aslamuqaili@yahoo. co. uk

ABSTRACT The transport sector consumes about 34% of the total commercial energy consumption in Pakistan. About 97% of fuel used in this sector is oil and the remaining 3% is CNG and electricity. The indigenous reserves of oil and gas are limited and the country is heavily dependent on the import of oil. The oil import bill is serious strain on the country's economy. The production, transportation and consumption of fossil fuels also degrade the environment. Therefore, it is important to explore the opportunities for clean renewable energy for long-term energy supply in the transport sector. Sindh, the second largest province of Pakistan, has about 250 km long coastline. The estimated average annual wind speed at 50 m height at almost all sites is about 6-7 m/s, indicating that Sindh has the potential to effectively utilize wind energy source for power generation and hydrogen production. A system consisting of wind turbines coupled with electrolyzers is a promising design to produce hydrogen. This paper presents an assessment of the potential of hydrogen production from wind energy in the costal area of Sindh, Pakistan. The estimated technical potential of wind power is 386 TWh per year. If the wind electricity is used to power electrolyzers, 347.4 TWh hydrogen can be produced annually, which is about 1.2 times the total energy consumption in the transport sector of Pakistan in 2005. The substitution of oil with renewable hydrogen is essential to increase energy independence, improve domestic economies, and reduce greenhouse gas and other harmful emissions.

Keywords: Hydrogen production; Wind energy; Transport sector; Pakistan *Corresponding author. Tel: +92-222771381; Fax: +92-221-771382

61

PARALLEL SESSION 4A: FUSION I TR0700293

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NUCLEAR CHALLENGES AND PROGRESS IN DESIGNING STELLARATOR POWER PLANTS

Laila El-Guebaly University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706, USA E-mail: [email protected]

ABSTRACT As an alternate to the mainline magnetic fusion , the stellarator concept offers a steady state operation without external driven current, eliminating the risk of plasma isruptions. Over the past 2-3 decades, stellarator power plants have been studied in the U.S., Japan, and Europe to enhance the physics and engineering aspects and optimize the design parameters that are subject to numerous constraints. The earlier 1980's studies delivered large with an average major radius exceeding 20 m. The most recent development of the compact stellarator concept has led to the construction of the National Compact Stellarator Experiment (NCSX) in the U.S. and the 3 years power plant study of ARIES-CS, a compact stellarator with 7.75 m average major radius, approaching that of tokamaks. The ARIES-CS first wall configuration deviates from the standard practice of uniform toroidal shape in order to achieve compactness. Modeling such a complex geometry for 3-D nuclear analysis was a challenging engineering task. A novel approach based on coupling the CAD model with the MCNP Monte Carlo code was developed to model, for the first time ever, the complex stellarator geometry for nuclear assessments. The most important parameter that determines the stellarator size and cost is the minimum distance between the plasma boundary and mid-coil. Accommodating the breeding blanket and necessary shield to protect the superconducting magnet represented another challenging task. An innovative approach utilizing a non-uniform blanket combined with a highly efficient WC shield for this highly constrained area reduced the radial standoff (and machine size and cost) by 25- 30%, which is significant. As stellarators generate more radwaste than tokamaks, managing ARIES-CS active materials during operation and after plant decommissioning was essential for the environmental attractiveness of the machine. The geological disposal option could be replaced with more attractive scenarios, such as recycling (within the nuclear industry) and clearance (or unconditional release to the commercial market). The ARIES-CS bioshield, cryostat, and individual magnet constituents qualify for clearance, representing -80% of the total waste volume. We developed a recycling approach for the non-clearable, in-vessel components using a combination of conventional and advanced remote handling equipment that can handle high doses of 3000 Sv/h or more. Several additional nuclear-related tasks received considerable attention during the ARIES-CS design process. These include the radial build definition, the well-optimized in-vessel components that satisfy the top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, and the overarching safety constraints to deliver a safe and reliable power plant. This paper provides a brief historical overview of the progress in designing stellarator power plants and a perspective to the successful integration of the nuclear activity into the final ARIES-CS design.

64 _ _.„, maum (lala ll(u I TR0700294

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SELF-ORGANIZED IGNITION OF A TOKAMAK PLASMA

Klaus Schoepf Institute for Theoretical Physics, University of Innsbruck, Austria E-mail: [email protected]

ABSTRACT The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product HTET, where XE represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual deposition [4,5] and its effect of reducing %E are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself- towards an ignited stable equilibrium state, even if the initial state is outside the positive power balance region. Since fuel is assumed to be constantly supplied for that, cold fuelling is the sole required injection to the system during the ignition dynamics. The known parameter discrepancy between stable and unstable ignited operation is found to be controllable by the fuelling rate, and thus may be reduced for a D-T fusion plasma to only a very few keV difference in the plasma temperature T. Further, the minimum heating path toward stable ignition was inspected; however, it was identified as an impracticable operational scenario due to substantial oscillations of the plasma density and temperature. Avoiding the latter, a heating regime with optimized low energy expense is proposed that leads conveniently to the stable ignition point at reasonable plasma temperatures in the 20 keV range. To minimize the energy required to heat the plasma beyond a state, from where it evolves by itself to ignited stable equilibrium, a relatively high power neutral beam applied for a short duration is found to be more efficient than a less powerful beam over a longer time period. References [1] IAEA Workshop on Burning Plasma Physics and Simulation, Tarragona, Spain, July 4-5, 2005; Workshop Summary in Fusion Sci. Techn. 49, 79 (2006) [2] K. Schoepf, T. Hladschik, Ann. Nucl. Energy 23, 59 (1996) [3] D. Anderson et al., Fusion Technology 23, 5 (1993) [4] K. Schoepf et al., Kerntechnik 67, 285 (2002) [5] K. Schoepf et al., TH/P6-3 at 21st IAEA Fusion Energy Conf, Chengdu, China, Oct. 2006

65 TR0700295

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

THE STUDY OF INERTIAL FUSION ENERGY PROBLEM VIA THE EQUATION OF STATE

Shalom Eliezer1'2'3, Jose M. Martinez Val2 and M. Murakami3 1. Soreq, NRC, Yavne, Israel 2. ETSII, Polytechnic University Madrid, Spain 3. ILE, Osaka University, Osaka, Japan E-mail: [email protected]

ABSTRACT It is known that many important physical phenomena can be obtained by analyzing the equation of state (EOS) of the stars1. For example, one can use the virial theorem and an ideal EOS to analyze the stars in a gravitational field. In this case, it is concluded that the star is unstable if D < 4/3 and it is stable if D > 4/3, where D is the ratio of the heat capacities at constant pressure and constant value. Furthermore, while a stable star contracts its internal energy increases and it gets hotter. At the same time it radiates energy. For 0= 5/3, half of the potential energy decrease is used to heat the star and the other half is irradiated. As can be deducted from this simple example, one can get a lot of insight into the study of the stars through the EOS. As is well known, a major breakthrough in inertial confinement fusion (ICF) occurred with the publication of J. Nuckolls et al.2 "Laser compression of matter to super-high densities: Thermonuclear applications". This important idea can be easily understood through EOS. Using for example the Thomas Fermi EOS for the deuterium-tritium nuclear fuel, it is concluded that it is energetically "cheaper" to compress the fuel rather than to heat it. On the other hand, it is known that the nuclear reaction rate is proportional to the density square. Therefore, the fusion gain G (= output energy/input energy) is significantly larger by compressing the full target while heating only a small portion of it. These schemes are known as spark ignition and fast ignition. The purpose of the target and driver designs in ICF is to obtain an appropriate fuel areal density (DR) and temperature (T) in order to achieve nuclear ignition and high gain. For a variety of different ICF designs: (a) spark ignition, (b) volume ignition, (c) fast ignition with picosecond lasers or (d) impact fast ignition, one requires different domains of initial DR and T values. Therefore the input energy for every scheme is in a domain set by the EOS and the mass of the fuel. Furthermore, the output energy is also a function of DR and T and therefore its value is deduced from the EOS. Another fundamental point in the compressed target evolution is radiation leakage. Radiation mean-free path scales at T7/2, which means that this energy sink term can be become rather small in highly compressed targets (high areal density) and moderate T.The paper deals with an analysis of the gain in ICF using the virial theorem and the Thomas Fermi EOS. Furthermore, simulation results of the recent impact fast ignition model3 will be presented. In this context we shall analyze for example, what could be better inertial fusion energy (IFE) choice: to have a compression of 200 g/cm3 and a temperature Ti or a compression of 600 g/cm3 and a temperature T2, where T2 is smaller than Ti. The answer depends on EOS. We analyze the EOS and suggest the optimum IFE design in order to achieve a maximum gain.

1. S. Eliezer, A. Ghatak, H. Hora and E. Teller, Fundemantal of Equations of State, Word Scientific, Singapore (2002). 2. J. Nuckolls et al., Nature 239, 139 (1972)3. Murakami et al., Nuclear Fusion 46, 99,2006.

66 TR0700296

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TRIANGULARITY EFFECTS ON THE COLLISIONAL DIFFUSION FOR ELLIPTIC TOKAMAK PLASMA

P. Martin and E. Castro Universidad Simon Bolivar, Dpto. Fisica, Apartado 89000, Caracas 1080A, Venezuela. E-mail: [email protected]

ABSTRACT In this conference the effect of ellipticity and triangularity will be analyzed for axisymmetric tokamak in the collisional regime. Analytic forms for the magnetic field cross sections are taken from those derived recently by other authors [1,2]. Analytical results can be obtained in elliptic plasmas with triangularity by using an special system of tokamak coordinates recently published [3-5]. Our results show that triangularities smaller than 0.6, increases confinement for ellipticities in the range 1.2 to 2. This behavior happens for negative and positive triangularities; however this effect is stronger for positive than for negative triangularities. The maximum diffusion velocity is not obtained for zero triangularity, but for small negative triangularities. Ellipticity is also very important in confinement, but the effect of triangularity seems to be more important. High electric inductive field increases confinement, though this field is difficult to modify once the tokamak has been built. The analytic form of the current produced by this field is like that of a weak Ware pinch with an additional factor, which weakens the effect by an order of magnitude. The dependence of the triangularity effect with the Shafranov shift is also analyzed. References 1. - L. L. Lao, S. P. Hirshman, and R. M. Wieland, Phys. Fluids 24, 1431 (1981) 2. - G. O. Ludwig, Plasma Physics Controlled Fusion 37, 633 (1995) 3. - P. Martin, Phys. Plasmas 7, 2915 (2000) 4. - P. Martin, M. G. Haines and E. Castro, Phys. Plasmas 12, 082506 (2005) 5. - P. Martin, E. Castro and M. G. Haines, Phys. Plasmas 12, 102505 (2005)

67 TR0700297

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

INNOVATIVE THREE-DIMENSIONAL NEUTRONICS ANALYSES DIRECTLY COUPLED WITH CAD MODELS OF GEOMETRICALLY COMPLEX FUSION SYSTEMS

Mohamed Sawan, P. Wilson, T. Tautges*, L. El-Guebaly, D. Henderson, G. Sviatoslavsky, T. Bohm, B. Kiedrowski, A. Ibrahim, B. Smith, R. Slaybaugh University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706, USA *Argonne National Laboratory, Argonne, IL 60439, USA E-mail: [email protected]

ABSTRACT Fusion systems are, in general, geometrically complex requiring detailed three-dimensional (3-D) nuclear analysis. This analysis is required to address tritium self-sufficiency, nuclear heating, radiation damage, shielding, and radiation streaming issues. To facilitate such calculations, we developed an innovative computational tool that is based on the continuous energy Monte Carlo code MCNP and permits the direct use of CAD-based solid models in the ray-tracing. This allows performing the neutronics calculations in a model that preserves the geometrical details without any simplification, eliminates possible human error in modeling the geometry for MCNP, and allows faster design iterations. In addition to improving the workflow for simulating complex 3- D geometries, it allows a richer representation of the geometry compared to the standard 2nd order polynomial representation. This newly developed tool has been successfully tested for a detailed 40° sector benchmark of the International Thermonuclear Experimental Reactor (ITER). The calculations included determining the poloidal variation of the neutron wall loading, flux and nuclear heating in the divertor components, nuclear heating in toroidal field coils, and radiation streaming in the mid-plane port.

The tool has been applied to perform 3-D nuclear analysis for several fusion designs including the ARIES Compact Stellarator (ARIES-CS), the High Average Power Laser (HAPL) inertial fusion power plant, and ITER first wall/shield (FWS) modules. The ARIES-CS stellarator has a first wall shape and a plasma profile that varies toroidally within each field period compared to the uniform toroidal shape in tokamaks. Such variation cannot be modeled analytically in the standard MCNP code. The impact of the complex helical geometry and the non-uniform blanket and divertor on the overall tritium breeding ratio and total nuclear heating was determined. In addition, we calculated the neutron wall loading variation in both the poloidal and toroidal directions. The final optics system of the HAPL power plant includes several metallic and dielectric mirrors that are sensitive to radiation. Although some of these mirrors are not in the direct line-of-sight of the neutron source, radiation scattering and streaming through the laser beam ports requires an assessment of the nuclear environment at the final optics to predict their lifetime. Detailed CAD models of the ITER FWS modules were analyzed to produce high resolution maps of nuclear heating, radiation damage and helium production. These clearly show the impact of the design heterogeneity details with the many coolant channels embedded in the module. In addition, hot spots produced in the vacuum vessel behind the module as a result of streaming through these coolant channels were evaluated. These examples will be presented to demonstrate the applicability of the tool to nuclear analysis of complex fusion systems.

68 PARALLEL SESSION 4B: FISSION REACTORS I TR0700298 I 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NAVAL APPLICATION OF BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM

Nam H. Kim, Tae W. Kim, Hyoung M. Son, Kune Y. Suh* Seoul National University, San 56-1 Sillim-dong, Gwanak-gu, Seoul, 151-744, Korea E-mail: [email protected]

ABSTRACT Past civilian N.S. Savanna (80 MWth), Otto-Hahn (38 MWth) and Mutsu (36 MW,h) experienced stable operations under various sea conditions to prove that the reactors were stable and suitable for ship power source. Russian nuclear icebreakers such as Lenin (90 MWthX 2), Arukuchika (150 MWth x 2) showed stable operations under severe conditions during navigation on the Arctic Sea. These reactor systems, however, should be made even more efficient, compact, safe and long- life, because adding support from the land may not be available on the sea. In order to meet these requirements, a compact, simple, safe and innovative integral system named Naval Application Vessel Integral System (NAVIS) is being designed with such novel concepts as a primary liquid metal coolant, a secondary supercritical carbon dioxide (SCO2) coolant, emergency reactor cooling system, safety containment and so on. NAVIS is powered by Battery Optimized Reactor Integral System (BORIS). An ultra-small, ultra-long-life, versatile-purpose, fast-spectrum reactor named BORIS is being developed for a multi-purpose application such as naval power source, electric power generation in remote areas, seawater desalination, and district heating. NAVIS aims to satisfy special environment on the sea with BORIS using the lead (Pb) coolant in the primary system. NAVIS improves the economical efficiency resorting to the SCO2 Brayton cycle for the secondary system. BORIS is operated by natural circulation of Pb without needing pumps. The reactor power is autonomously controlled by load-following operation without an active reactivity control system, whereas B4C based shutdown control rod is equipped for an emergency condition. SCO2 promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the SCO2 Brayton cycle efficiency as high as 45 % furnishes small sized nuclear reactors with economical benefits on the plant construction and maintenance. BORIS is being designed to generate 23 MWj, for at least twenty consecutive years without refueling and to meet the naval nuclear system goals of compactness, safety, reliability and economics. BORIS utilizes proliferation-resistant nitride fuel with a high thermal conductivity and open cartridge type core without individual subassemblies. BORIS consists of a reactor module, heat exchangers, coolant module, guard vessel, ERCS, secondary system, and safety containment.

70 TR0700299

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

ENHANCING VVER ANNULAR PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

Gray S. Chang Idaho National Laboratory: Idaho Falls, ID 83415 USA E-mail: [email protected]

ABSTRACT Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near- term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241 Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems.

71 TR0700300

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PRELIMINARY DESIGN of SMART FUEL

Yonghwan Kim, Dongguen Ha, Sungkyu Park, Keeil Nahm, Kyuseok Lee, Jungha Kim Korea Nuclear Fuel Co. Ltd, 493 Duck-Jin Dong Dae-Jeon City, Korea E-mail: [email protected]

ABSTRACT SMART (System-integrated Modular Advanced ReacTor) is a novel light water rector with a modular, integral primary system configuration. This concept has been developing a 660MWt by Korean Nuclear Power Industry Group with KAERI. SMART is being developed for use as an energy source for small-scale power generation and seawater desalination. Although the design of SMART is based on the current pressurized water reactor technology, new technologies such as enhanced safety, and passive safety have been applied, and system simplification and modularization, innovations in manufacturing and installation technologies have been implemented culminating in a design that has enhanced safety and economy, and is environment -friendly.

In this paper described the preliminary design of the nuclear FUEL for this SMART, the design concept and the characteristics of SMART FUEL. In specially this paper describe the optimization of grid span adjustment to improve the thermal performance of the SMART FUEL as well as to improve the seismic resistance performance of the SMART FUEL, it is not easy to improve the both performance simultaneously because of design parameter of each performance inversely proportional. SMART FUEL enable to extra-long extended fuel cycle length and resistance of proliferation, enhanced safety, improved economics and reduced nuclear waste.

72 TR0700301

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

SOLITARY BURN-UP WAVE SOLUTION IN MULTI-GROUP DIFFUSION-BURNUP COUPLED SYSTEM

X.-N. Chen andW.Maschek i Institute for Nuclear and Energy Technologies, FZK, Germany, email: E-mail: xue-nong. [email protected]. de 2 Institute for Nuclear and Energy Technologies, FZK, Germany, E-mail: werner. [email protected]. de

ABSTRACT The solitary burn-up wave solution or CANDLE concept becomes well-known in the series of ICENES meetings [1, 2, 3, 4, 51. The basic idea behind this concept is the existence of self- 238 232 propagating nuclear breeding/burning solitary wave in a fertile medium of U or Th. Natural thorium and uranium fuel can be used for this concept and a high burn-up can be achieved. Therefore no fuel enrichment and reprocessing are needed and a long continuous operation duration is possible. Fundamental insight into this new type of reactor was given by van Dam (2000) [6] and Seifritz (2000) [7]. An exact solitary wave solution was found in a 1-D single group diffusion equation with suitably assumed burn-up dependent coefficients and a nonlinear term of feedback effects [6]. The solitary wave solution was found analytically as well in the same diffusion equation without feedback effects but coupled by simple burn-up equations [7], where the solitary wave is generally fore-aft asymmetric (skew). Intensive numerical studies have been made by Sekimoto, Ryu and Yoshimura (2001) [8]. They considered that a nuclear ignition region charged with plutonium or enriched uranium was set at an end of the core and natural or possibly depleted uranium was charged in the remaining region. They solved the multi-group diffusion and burn-up equations numerically and demonstrated the feasibility of this new concept. The main purpose of this paper is to achieve analytical solutions with a multi-group model for this kind of problem. A neutronic model, i.e. two-group diffusion equations coupled with burn-up equations, are proposed for obtaining a 1-D asymptotic solution in a moving 238 239 240 241 242 coordinate system. In the burn-up equations, only U, Pu, Pu, Pu, Pu, a typical fission product pair (FPP) and a burnable poison (BP) are considered. Radioactive decay processes are neglected, because the radioactive decay processes are either too short or too long with respect to the considered time scale of the order of several years. Hence, as the results of the solution of burn-up equations, macroscopic cross sections are only functions of the neutron fluences and the initial values of atom number density and, consequently, the diffusion equations become nonlinear differential-integral ones. Although this multi-group coupled model looks much more complicated, at least in the 1-D case it is still analytically solvable like in the single-group model! The existence conditions of solitary wave solution will be investigated, although they are much more complicated than in the single group case. The interaction mechanism of thermal and fast modes within the solitary wave solution can be revealed. Moreover, this solution can be applied in a thermal-fast spectrum hybrid core, e.g., a . References [1] E. Teller, M. Ishikawa and L. Wood, Proceedings ICENES'1996 Obninsk, Russia (1996) 151. [2] H. Sekimoto and K. Ryu, Proceedings ICENES'2000 Petten, Netherlands (2000).

73 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

[3] H. van Dam, Proceedings ICENES'2000 Petten, Netherlands (2000). [4] X.-N. Chen, E. Kiefhaber and W. Maschek, Proceedings ICENES'2005 Brussels, Belgium (2005). [5] H. Sekimoto and S. Miyashita, Proceedings ICENES'2005 Brussels, Belgium (2005). [6] H. van Dam, Annals of Nuclear Energy 27 (2000) 1505. [7] W. Seifritz, Kerntechnik 65 (2000) 261. [8] H. Sekimoto, K. Ryu and Y. Yoshimura, Nuclear Science and Technology 139 (2001) 306. [9] X.-N. Chen and W. Maschek, Annals of Nuclear Energy 32 (2005) 1377.

74 TR0700302

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

WHOLE CORE TRANSPORT CALCULATION FOR THE VHTR HEXAGONAL CORE

Jin-Young Cho*, Kang-Seog Kim*, Chung-Chan Lee*, and Han-Gyu Joo** *Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353, Korea, Fax: +82 42 868 8990, E-mail: jyoung@kaeri. re. kr **Department of Nuclear Engineering, Seoul National University, San 56-1, Sillim-dong, Kwanak-gu, Seoul, 151-742, Korea, Fax: +82 2 889 2688, E-mail: joohan@snu. ac.kr

ABSTRACT Recently, the DeCART code which performs the whole core calculation by coupling the radial MOC transport kernel with the axial nodal kernel has equipped a kernel to deal with the hexagonal geometry and applied to the VHTR hexagonal core to examine the accuracy and the computational efficiency of the implemented kernel. The implementation includes a modular ray tracing module based on the hexagonal assembly and a multi-group CMFD module to perform an efficient transport calculation. The requirements for the modular ray are: (1) the assembly based path linking and (2) the complete reflection capabilities. The first requirement is met by adjusting the azimuthal angle and the ray spacing for the modular ray to construct a core ray by the path linking. The second requirement is met by expanding the constructed azimuthal angle in the range of [0,30°] to the remained range to reflect completely at the core boundaries. The considered reflecting surface angles for the complete reflection are 30n's (n=l,2,...,12). The CMFD module performs the equivalent diffusion calculation to the radial MOC transport calculation based on the homogenized structure units. The structure units include the hexagonal pin cells and gap cells appearing at the assembly boundary. Therefore, the CMFD module is programmed to deal with the unstructured cells such as the gap cells. The CMFD equation consists of the two parts of (1) the conventional FDM and (2) the current corrective parts. Since the second part of the CMFD equation guarantees the reproducibility of the radial MOC transport solutions for the cell averaged reaction rate and the net current at the cell surfaces, how to build the first part of the CMFD equation is not important. Therefore, the first part of the CMFD equation is roughly built by using the normal distance from the gravity center to the surface.

The VHTR core uses helium as a coolant which is realized as a void hole in a neutronics calculation. This void hole which occupies several cells has no influence on the MOC transport calculation, but it introduces a near singular matrix for a CMFD formulation. This void problem is resolved by introducing a lumped CMR scheme in which the void cells are collapsed to one equation to produce one rebalancing factor. The lumped CMR scheme causes the original CMFD equation not to converge. Therefore, the CMFD calculation is stopped if the residual error of the CMFD solution does not reduce, and return to the radial MOC transport calculation.In the comparison of the computational result with the MCNP code for the VHTR 2-D core problem, DeCART shows about 200 ~ 500 pcm difference in the eigenvalue, and less than 1.0 % difference in the assembly power distribution. For the computing time, DeCART takes less than 5 hours on a PENTIUM-IV 3.0 GHz personal computer. Those results indicated that the hexagonal module of the DeCART code worked very well within an affordable computing time.

75

PARALLEL SESSION 4C: SOLAR ENERGY II TR0700303

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

MICROSCOPIC MORPHOLOGICAL COMPENSATION FOR PHASE-SEPARATED COMPOSITE FILM THICKNESS

Chia -Fu, Chang1, Yi-Ci, Chan2, Zou-ni, Wan3 Kun Shan University of Technology No.661,duanl,Jhong Shan Rd,Yilan City,260,Taiwan(R.O.C) E-mail: [email protected]

ABSTRACT The generic structure of our bimeso gens is shown in and for a typical blue-phase mixture of the type we describehere we usemixtures of the ratio 33.4% {n =2.6), 34.1% in =6.57), 36.6% in = 11.15) with of the high twisted power (HTP) agent BD HI 381 (available from Merck Chemicals and described in ref. We thenstudied theelectric-field dependency of the selective reflection in BP I* at 20.7 °C by applying increasing and the decreasing pulsed alternating current (a.c.) electric fields (100 Hz).

78 TR0700304

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

A NOVEL GUI MODELED FUZZY LOGIC CONTROLLER FOR A SOLAR POWERED ENERGY UTILIZATION SCHEME

I. H. Altas1,* and A. M. Sharaf2 and 1>*Dept. of Electrical and Electronics Engineering, Karadeniz Technical University, Trabzon, Türkiye E-mail: [email protected] 2Dept. of Electrical and Computer Engineering, University of New Brunswick, Canada *: Currently a visiting scholar at the University of New Brunswick, Canada E-mail: [email protected]

ABSTRACT Photovoltaic PVA-solar powered electrical systems comprise different components and subsystems to be controlled separately. Since the generated solar power is dependant on uncontrollable environmental conditions, it requires extra caution to design controllers that handle unpredictable events and maintain efficient load matching power. In this study, a photovoltaic (PV) solar array model is developed for Matlab/Simulink GUI environment and controlled using a fuzzy logic controller (FLC), which is also developed for GUI environment. The FLC is also used to control the DC load bus voltage at constant value as well as controlling the speed of a PMDC motor as one of the loads being fed. The FLC controller designed using the Matlab/Simuling GUI environment has flexible design criteria's so that it can easily be modified and extended for controlling different systems. The proposed FLC is used in three different parts of the PVA stand alone utilization scheme here. One of these parts is the speed control of the PMDC load, one of the other parts is controlling the DC load bus voltage, and the third part is the maximum power point (MPPT) tracking control, which is used to operate the PVA at its available maximum power as the solar insolation and ambient temperature change.

This paper presents a study of a standalone Photovoltaic energy utilization system feeding a DC and AC hybrid electric load and is fully controlled by a novel and simple on-line fuzzy logic based dynamic search, detection and tracking controller that ensures maximum power point operation under excursions in Solar Insolation, Ambient temperature and electric load variations. The maximum power point MPP-Search and Detection algorithm is fully dynamic in nature and operates without any required direct measurement or forecasted PV array information about the irradiation and temperature. An added Search sensitivity measure is defined and also used in the MPP search algorithm to sense and dynamic response for other reduced MPP operating points. The proposed dual action control scheme is very effective for Large PV Array installations using only Common PVA current and voltage signals are measured. The paper investigates a novel online dual-action fuzzy logic control scheme for maximum power point tracking. The main part of this novel dual-action FL controller has a similar structure as that of classical FL controllers but with a different choice of input signals. The proposed novel dynamic FLC controller uses the power error and the ratio AP/AI as the two input signals instead of using error signal and its change over one sampling period. Therefore the rule generation philosophy here differs from that of a usual FLC structure. The auxiliary part of the dual fuzzy MPP tracking controller is introduced as a novel approach to handle the dead zones left from the main part. The proposed MPP detection algorithm and the dual fuzzy logic MPP tracking controller are validated using the Matlab/Simulink software environment by digitally simulating the PV array scheme feeding

79 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye hybrid DC and AC loads. Besides the MPP detector and dual fuzzy logic MPP tracking controller, the scheme includes three more control units, one of them is for the voltage control of the common dc load bus, the second one is for voltage and frequency control of AC load bus, and the third one is for the speed control of the motor type loads in both DC and AC sides.

80 TR0700305

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

DESIGN AND PERFORMANCE OF GAN BETAVOLTAIC DEVICE

Hyun-Kyu Jung1, Nam-Ho Lee1, Sang-Kwon Lee2 1 Korea Atomic Energy Research Institute (KAE(KAERI)R , PO Box 105, Yusong, Daejon 305-600, Korea 2 Chunbuk National1 University Univei , Jeonju, Korea E-mail: [email protected]

ABSTRACT Semiconductor betavoltaic cell employs a semiconductor pn junction and a radioisotope that emits beta particles. This device can be used for remote applications requiring a long life power and for minimizing the size. In order to obtain a useful beta radiation power from the nuclear waste, Nickel-63 and Carbon-14 radioisotopes were selected. Ni-63 energy source adapted to a wideband GaN p-n junction as well as silicon device but C-14 has the higher energy which affects the extensive degradation of performance in silicon device and can adapt only to a GaN semiconductor showing the radiation tolerance. The experimental results for I- V characteristics were compared and analyzed. This paper also presents the design concepts for high efficiency GaN betavoltaic cell.

81 TR0700306 !

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SOLAR CONTROL ON IRRADIATED TA2OS THIN FILM

Nilgün DOĞAN BA YDOĞAN3, Esra ÖZKAN ZAYİMb * a* Institute of Energy, bScience and Literature Faculty Istanbul Technical University, Ayazaga Campus, Maslak, 34469, Istanbul, Türkiye E-maila: [email protected], E-mailb: [email protected]

ABSTRACT Thin films consisting of T^O^ have been used in industry in applications related to thin-film capacitors, optical waveguides, and antireflection coatings on solar cells. Ta2Û5 films are used for several special applications as highly refractive material and show different optical properties depending on the deposition methods. Sol-gel technique has been used for the preparation of Ta2Û5 thin films. Ta2Û5 thin films were prepared by sol-gel prosses on glass substrates to obtain good quality films. These films were exposed to gamma radiation from Co-60 radioisotope. Ta2Û5 coated thin films were placed against the source and irradiated for 8 different gamma doses; between 0.35 and 21.00 kGy at room temperature. Energetic gamma ray can affect the samples and change its colour. On the other hand some of the Ta2Û5 coated thin films were irradiated with beta radiation from Sr-90 radioisotope. The effect of gamma irradiation on the solar properties of Ta2Û5 films is compared with that of beta irradiation. The solar properties of the irradiated thin films differ significantly from those of the unirradiated ones. After the irradiation of the samples transmittance and reflectance are measured for solar light between 300 and 2100 nm, by using Perkin Elmer Lambda 9 UV/VIS/NIR Spectrophotometer. Change in the direct solar transmittance, reflectance and absorptance with absorbed dose are determined. Using the optical properties, the redistribution of the absorbed component of the solar radiation and the shading coefficient (SC) are calculated as a function of the convective heat-transfer coefficient. Solar parameters are important for the determination of the shading coefficient. When the secondary internal heat transfer factor (

Corresponding Author Tel: +90.0212.285.30.09

82 TR0700307

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

CONCEPTUAL DESIGN OF GAN BETAVOLTAIC BATTERY USING IN CARDIAC PACEMAKER

M. Mohamadian, S. A. H. Feghhi1, H. Afarideh Faculty of Physics and Nuclear Sciences, Amirkabir University of Technology, Tehran, Iran E-mail: [email protected]

ABSTRACT Introduction: Pacemaker is an electronic biomedical device which stimulates and regulates or amplify the human heartbeat by delivering weak electrical pulses to the cardiac muscle at regular intervals when its natural regulating mechanisms break down. Developments in design and implementation of power source in adjacent to advances in electronic circuitry is an important aspect in optimization of pacemakers. For instance, many implant patients continue to outlive their batteries and require costly and risky replacement surgery. So such device needs to have high energy density power source and maintain a stable current and voltage for a long period of time to avoid frequent replacements. In addition, the size is also an important consideration for implantable batteries. Betavoltaic batteries are being researched as a suitable source for these applications. Also, these batteries have vast application in which the replacement of batteries is highly inconvenient, such as in oil and mining industries, which often place sensors in dangerous or hard-to-reach locations. The purpose of the present investigation is determination of the optimal parameters of low energy GaN betavoltaic battery in artificial cardiac pacemakers using MCNP code which have higher efficiency than those available with previous devices, especially thermoelectric converters (-15%).

Material and Methods: In this design, two p-n diode structures from GaN semiconductor were used to collect the charge from a layer of 63Ni as a source which is centered between the two p-n junctions. MCNP simulation results have been used to determine the amount of electron current from interaction of beta particles in p-n junctions.

Results and Discussion: Calculation results indicate that the short circuit current, open circuit voltage and efficiency of a single device are 1.1 uA/cm2, 2.7 volt and 25%, respectively. Also, it's concluded that with suitable arrangement of these single devices, one could construct a battery with required current, voltage and power for this application. So betavoltaic batteries can be useful in low-power equipments, especially due to more safety because of very short ranges of beta particles emitted form using radioisotope which require minimal shielding and are unable to penetrate human skin.

Keywords: Betavoltaic battery, Cardiac pacemaker

Corresponding author; Tel.: +98 21 64542591 & +98 9122109461 Fax:+98 21 66495519

83

PARALLEL SESSION 5A: ACCELERATOR DRIVEN SYSTEMS II TR0700308

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

R&D ACTIVITIES AROUND THE EUROTRANS ACCELERATOR FOR ADS APPLICATIONS

Jean-Luc BIARROTTE and Alex C. MUELLER CNRS/IN2P3, IPN Orsay On behalf of the EUROTRANS WP1.3 collaboration, France E-mail: [email protected]

ABSTRACT An Accelerator Driven System (ADS) for transmutation of nuclear waste typically requires a 600 MeV - 1 GeV accelerator delivering a proton flux of a few mAs for demonstrators, and of a few tens of mAs for large industrial systems. Such a machine belongs to the category of the high- power proton accelerators, with an additional requirement for exceptional "reliability": because of the induced thermal stress to the subcritical core, the number of unwanted "beam-trips" should not exceed a few per year, a specification that is several orders of magnitude above usual performance. This paper briefly describes the reference solution adopted for such a machine, based on a linear superconducting accelerator, and presents the status of the R&D performed in this context. This work is performed within the 6th Framework Program EC project "EUROTRANS"2.

2 EC Contract N° FI6W 516520, "EUROTRANS"

86 TR0700309

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

NEUTRONICS AND SHIELDING ISSUES OF ADS

Hamid A. Abderrahim3, T. Aoust8, E. Gonzalez5, W. Haeck8, E. Malambu3, J.M. Martinez-Valc, Y. Romanetsd, G. Van den Eyndea, P. Vazd'3, C. Vicenteb aSCK»CEN, Boeretang 200, B-2400 Mol, Belgium bCIEMAT, Avda. Complutense, 22, E-28040 Madrid, Spain CUPM/ETSII, Jose Gutierrez Abascal, 2, E-28006 Madrid, Spain dITN, Estrada Nacional 10, P-2686-953 Sacavem, Portugal E-mail: [email protected]

ABSTRACT Accelerator Driven Systems (ADS) are hybrid systems consisting of a high-intensity proton accelerator with beam energy in the hundreds of MeV range impinging in a target of a heavy element and coupled to a sub-critical core. The intense (of the order of 1015 n/cm2/s) and fast neutron fluxes produced by the spallation reactions triggered by the impinging protons in the target can be used to induce fission reactions in the actinides and capture reactions in the long- lived fission products in the fuel assemblies in the core of the system. ADS have been considered during the last fifteen years as one of the promising technological solutions for the transmutation of nuclear waste, reducing the radiotoxicity of the high-level nuclear waste and reducing the burden to the geological repositories. The European Commission's Green Paper entitled "Towards a European Strategy for the Security of Energy Supply" clearly pointed out the importance of nuclear energy in Europe. With 145 operating reactors producing a total power of 125 GWe, the resulting energy generation of 850 TWh per year provides 35% of the electricity consumption of the European Union. The Green Paper also points out that the nuclear industry has mastered the entire with the exception of waste management and for this reason, "focusing on waste management has to be continued". Amongst the several solutions being studied in recent years, MYRRHA (concept developed at SCK-CEN, Belgium), XADS (design studies co-funded by the European Union in the framework of the 5th Framework Programme) and XT-ADS and EFIT (acronyms standing for an experimental machine and for the long term transmuter to be deployed on an industrial scale, both in the EUROTRANS project of the 6th Framework Programme) have deserved the attention of different communities of specialists in the field of Nuclear Technology and Radioactive Waste Management. Although these machines have been designed with different parameters, their implementation and deployment have in common the fact that they raise cutting edge scientific and technological problems, associated to the operation of the high-intensity proton accelerator, the high-power (in the multi-MegaWatt range) delivered to the target and the material damage in the target and surrounding structures. The thermal power in the core, the thermal-hydraulic aspects associated to the heat removal in steady state and also in transient mode, the subcriticality level of the system and the efficiency of the transmutation process, is particularly sensitive to the core design (geometry, number of subassemblies, fuel composition, among many other aspects). Neutronic and shielding issues and the computation and mapping of neutron fluxes and doses are important throughout all stages of design of these systems. In this paper, i) the main characteristics and parameters of the ADS systems previously alluded to will be reviewed ii) the neutronics and

3 Corresponding author: [email protected]

87 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye shielding calculations of relevance for the design of the ADS systems, for radiation damage and for purposes will be extensively described.

88 TR0700310

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

THERMAL FEATURES OF SPALLATION WINDOW TARGETS

Jose M. Martinez-Val, F. Sordo, P. T. Leon E.T.S.Ingenieros Industrials 28006 Madrid (Spain) E-mail: [email protected]

ABSTRACT Subcritical nuclear reactors have been proposed for a number of applications, from energy production to fertile-to-fissile conversion, and to transmutation of long-lived radionuclei into stable or much shorter-lived nuclei. The main advantage of subcritical reactors is their large reactivity margin for not to attain prompt-supercritical power surges. On the contrary, subcritical reactors present some economic drawbacks and technical complexities that deserve suitable attention in the R&D phase. Namely, they need a very intense neutron source in order to keep the neutron flux and the reactor power at the required level. The most intense neutron source seems to be based on the proton-induced (or deuteron-induced) spallation reaction in heavy nuclei targets, which present very demanding thermal features that must be properly limited. Those limits pose upper bounds to the neutron yield of the target. In turn, the limits depend on the features of the impinging particle beam and the material composition and geometry of the target. Although the potential design window for spallation targets is rather wide, the analysis presented in this paper identifies specific topics that must properly be covered in the detailed project of a spallation source, in order to avoid unacceptable temperatures and mechanical stresses in the most critical parts of the source.

In this paper, some calculations are reported on solid targets (water cooled or helium cooled) and molten metals targets. It is seen that thermal-hydraulic and mechanical calculations of spallation targets are fundamental elements in the coherent design of this type of very intense neutron sources. This coherence implies the need of a suitable trade-off among the relevant beam parameters (proton energy, total intensity and cross-section shape) and the features of the target (structural materials, coolant characteristics and target geometry). The goal of maximizing the neutron yield has to be checked against the safety criteria regarding the source integrity, notably those depending on the thermal performance of the system.

89 TR0700311

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

CERMET FUEL BEHAVIOUR AND PROPERTIES IN ADS REACTORS

Didier Haas, A. Fernandez, D. Staicu, J. Somers Joint Research Centre Institute for Transuranium Elements, Germany E-mail: [email protected] W. Maschek, X. Chen ForschungsZentrum Karlsruhe, Germany

ABSTRACT Within the EUROTRANS Integrated Project co- financed within the 6th Framework Programme of the European commission, the sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. Within the EUROTRANS Programme, the domain AFTRA is responsible to develop and provide the data basis for the fuels to be used in the European Facility for Industrial Transmutation (EFIT). The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmuter fuels are strongly linked with the core design and safety parameters, the fuel properties and the ease of fabrication and reprocessing. This study concerns the behaviour and properties of fuels with molybdenum as inert matrix. The status of the development work was presented at the last ICENES conference [1]. Since then, the design of the European Facility for Industrial Transmutation (EFIT) was developed and the transmutation capability, the burn-up behaviour, the reactivity swing and power peaking factors, and the safety performance were determined for different cores with inert matrix fuels like MgO and Mo. For the EFIT, the CERMET with the Mo matrix is recommended as the reference fuel and CERCER with the MgO matrix as a back-up solution. The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were measured, and the thermal conductivity was deduced. The thermal conductivity of the CERMET fuels was also predicted using a model proposed in [1], with a microstructure corresponding to a random distribution of spheres, with overlapping. This model microstructure takes into account the negative effects arising from the possible formation of small agglomerates of inclusions in the CERMET fuels. The agreement between the theoretical and calculated values is relatively good (the error is between 0 and 15% of the value of the thermal conductivity).Irradiation programmes are in the final stage of preparation (and will start in 2007) to determine the in-reactor performance of the material. CERMET fuel pins are incorporated in two experiments: - Two pins will be loaded in the PHENIX reactor in Marcoule, within the FUTURIX FTA experiment [2]. These fuels have been fabricated at ITU in 2005-2006, according to the reference fabrication process in the Minor Actinide Laboratory, namely the infiltration of minor actinide solution in solid particles. These fuels have been fully characterised in terms of pellet structure, thermal properties, re-sintering behaviour, etc... The aim of the experiment is the investigation of the fuel behaviour under high fast neutron flux condition, and its comparison with other fuel types (CERCER, nitride and metallic).The completion of the irradiation is foreseen in 2009. - Two further CERMET fuel pins will be irradiated in the HFR reactor in Petten: the HELIOS experiment [3]. There the aim is the study of the gas (including Helium, produced by Am241 transmutation chain) production and release, in comparison with Am targets supported in a pure zirconia matrix. The post-irradiation examinations to be performed after 10 irradiation cycles will be concluded in 2009.

90 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

Safety studies for optimised EFIT core designs, developed within the AFTRA domain were performed. The safety coefficients and indicators were determined for each core, and various transients were investigated. For the new low power cores of EFIT with a power class of ~ 400 MWth and a fuel power density of- 250 MW/m3 it can be demonstrated that the CERMET cores behave favourably and the design limits of the fuels are not violated [4]. Results indicate that the T91 cladding used as clad more severely restricts possible design options. This report will present the status of the neutronic and safety studies for the EFIT core, the CERMET thermal properties determination results, as well as the final results of the fabrication, characterisation and irradiation conditions in these two new fuel irradiation experiments. References 1. Haas & al. ICENES Conference 2005 Brussels 2. Jaecki &al. GLOBAL 2005 3. Scaffidi-Argentina & al. ICAPP 2006 4. W. Maschek, X. Chen, C. Matzerath Boccaccini, A. Rineiski, J. Wallenius , Sobolev, P. Smith, R. Thetford, J.P. Ottaviani, S.Pillon, D. Haas, First Results of Safety Analyses for ADTs with CERCER and CERMET Fuels within the EUROTRANS-AFTRA Program, Ninth Information Exchange Meeting, Actiniae and Fission Product Partitioning & Transmutation, France, 25-29 September 2006

91 __™ um mai umiM 1 TR0700312

13lh International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ASSESSMENT OF THE TRANSMUTATION CAPABILITY OF AN ACCELERATOR DRIVEN SYSTEM COOLED BY LEAD BISMUTH EUTECTIC ALLOY

F. Bianchi(l), V. Peluso(l), R. Calabrese(l), X. Chen(2), W. Maschek(2) (1) ENEA - Italian National Agency for New Technologies, Energy and the Environment, Via Martiri di Monte Sole 4, 40129 Bologna (Italy) (2) FZK - Research Centre Karlsruhe, D-76021 Karlsruhe (Germany) E-mail: fosco. bianchi@bologna. enea. it

ABSTRACT 1. PURPOSE The reduction of long-lived fission products (LLFP) and minor actinides (MA) is a key point for the public acceptability and economy of nuclear energy. In principle, any nuclear fast reactor is able to burn and transmute MA, but the amount of MA content has to be limited a few percent, having unfavourable consequences on the coolant void reactivity, Doppler effect, and delayed neutron fraction, and therefore on the dynamic behaviour and control. Accelerator Driven Systems (ADS) are instead able to safely burn and/or transmute a large quantity of actinides and LLFP, as they do not rely on delayed neutrons for control or power change and the reactivity feedbacks have very little importance during accidents. Such systems are very innovative being based on the coupling of an accelerator with a subcritical system by means of a target system, where the neutronic source needed to maintain the neutron reaction chain is produced by spallation reactions. To this end the PDS-XADS (Preliminary Design Studies on an experimental Accelerator Driven System) project was funded by the European Community in the 5th Framework Program in order both to demonstrate the feasibility of the coupling between an accelerator and a sub-critical core loaded with standard MOX fuel and to investigate the transmutation capability in order to achieve values suitable for an Industrial Scale Transmuter. This paper summarizes and compares the results of neutronic calculations aimed at evauating the transmutation capability of cores cooled by Lead-Bismuth Eutectic alloy and loaded with assemblies based on (Pu, Am, Cm) oxide dispersed in a molybdenum metal (CERMET) or magnesia (CERCER) matrices. It also describes the constraints considered in the design of such cores and describes the thermo-mechanical behaviour of these innovative fuels along the cycle.

2. DESCRIPTION OF THE WORK The U-free composite fuels (CERMET and CERCER) were selected for this study, being considered at European level the most promising among the various technologies foreseen for designing ADS core with enhanced waste transmutation. The neutronic calculations were performed with a special ERANOS Procedure (MECONG) that utilizes a RZ core models for the description of the core geometry and represents the various regions in homogeneous manner. A multi-recycling scenario was hypothesized and a proper amount of plutonium and minor actinides was supplied at the beginning of each cycle in order to ensure the same operating reactivity (keff=0.97). Moreover some core design parameters were changed in order to investigate the capability of such cores to burn/transmute MA with acceptable safety features. The behaviour of fuels pin during the cycle in terms of fuel temperature, internal pressure, stresses and strains was investigated by using TRANSURANUS code.

92 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

3 RESULTS AND CONCLUSIONS The preliminary analysis shows that a good compromise between transmutation and core performance can be achieved for both fuels increasing the core power. Of course the increase of the core size has a significant implication on the overall plant architecture, in particular on accelerator and spallation module.

93

PARALLEL SESSION 5B: FISSION REACTORS II TR0700313

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

REACTOR PHYSICS SIMULATIONS WITH COUPLED MONTE CARLO CALCULATION AND COMPUTATIONAL FLUID DYNAMICS

Volkan Seker, Justin W. Thomas and Thomas J. Downar Purdue University Nuclear Engineering Building 400 Central Drive West Lafayette, IN, 47907, USA E-mail: [email protected]; [email protected]; [email protected]

ABSTRACT The interest in high fidelity modeling of nuclear reactor cores has increased over the last few years and has become computationally more feasible because of the dramatic improvements in processor speed and the availability of low cost parallel platforms. In the research here high fidelity, multi-physics analyses was performed by solving the neutron transport equation using Monte Carlo methods and by solving the thermal-hydraulics equations using computational fluid dynamics. A computation tool based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program "McSTAR" along with the verification and validation efforts. McSTAR is written in PERL programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STAR-CD for every region. Three different methods were investigated and two of them are implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. The necessary input file manipulation, data file generation, normalization and multi-processor calculation settings are all done through the program flow in McSTAR. Initial testing of the code was performed using a single pin cell and a 3X3 PWR pin- cell problem. The preliminary results of the single pin-cell problem are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k and the power profile was observed. The comparison of the results of 3X3 pin-cell problem is on-going. Furthermore, methods to enhance the convergence and accelerate the computation are being investigated.

96 TR0700314

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

FORMULATION OF DETECTOR RESPONSE FUNCTION TO CALCULATE THE POWER DENSITY PROFILES USING IN-CORE NEUTRON DETECTORS

ASM Sabbir Ahmed1, J. K. Shultis2, Joerg Peter1, Wolfrad Semmler1 'German Cancer Research Center (DKFZ), Heidelberg University, Im Neuenheimer Feld 280, ECT0204, D-69120, Germany department of Mechanical & Nuclear Engineering, Kansas State University, Manhattan, KS- 66506, USA Email: [email protected]

ABSTRACT By measuring neutron fluxes at different locations throughout a core, it's possible to derive the power-density profile P* (W cm"3), at an axial depth z of fuel rod k. Micro-pocket fission detectors (MPFD) have been fabricated to perform such in-core neutron flux measurements [1]. The purpose of this study is to develop a mathematical model to obtain axial power density distributions in the fuel rods from the in-core responses of the MPFDs.

97 TR0700315

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SCENARIO ANALYSIS FOR TRANSURANIC TRANSMUTATION BY USING FAST REACTORS

Chang Joon Jeong Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong, Daejeon, 305-600, Korea E-mail: [email protected]

ABSTRACT Symbiotic fast reactor scenarios with the existing nuclear power systems have been analyzed from the viewpoint of a transuranics transmutation. In this study, a sodium-cooled fast reactor (SFR) and an accelerator driven system (ADS) are considered as representative fast reactor systems. For a comparative analysis of the fuel cycle options, the once-through fuel cycle was at first analyzed based on the current construction plan and the currently operating nuclear power plants such as the pressurized water reactor (PWR) and the Canada deuterium uranium (CANDU) reactor. After setting up a once-through fuel cycle model, the SFR and ADS scenarios were modeled based on the same nuclear energy demand prediction used for the once-through fuel cycle. Then important fuel cycle parameters such as the amount of the spent fuel and corresponding plutonium, minor actinides and fission products inventories were estimated and compared with those of the once-through fuel cycle. In this fuel cycle model, the Pyro process is assumed for all the spent fuel recycling. In the process all the actinides are recovered and some fraction of the fission product is removed. The deployment fractions of the fast reactor are 25, 10 and 20% for the periods of 2030-2040, 2041-2070 and 2071-2100, respectively. In order to feed the fast reactor systems, it was also assumed that the PWR and CANDU spent fuels are reprocessed from 2025 and the fast reactor spent fuel reprocessing begins in 2035. The fuel cycle calculation was performed by the DYMOND code, which has been used for an analysis of the Generation-IV roadmap studies.

The analysis results of the once-through fuel cycle can be summarized as follows: - The nuclear power demand is expected to grow to 25.2 GWe in the year 2100. - The total spent fuel inventory is expected to be 650001 in 2100. - The transuranics and fission product inventories are estimated to be 660 and 2390 t, respectively, in 2100.

The fast reactor cycle analysis results can be summarized as follows: - The SFR and ADS can transmute the transuranics by 56 and 130 t, respectively, which correspond to a reduction of 8 and 20% when compared to the once-through cycle. - The total fission product inventories of the SFR and ADS cycles are 2800 and 2360 t, respectively, which are similar to that of the once-through cycle. For the long-lived fission products such as 129I and 99Tc, the SFR transmutes 129I and 99Tc by 0.8 and 3.2 t, respectively, while the ADS transmutes 129I and 99Tc by 4 and 16 t, respectively. From a transmutation point of view, the ADS is better than the SFR. The transmutation rate is not that high, but it can be improved by increasing the deployed capacity of the fast reactor in the scenario. In the future, we need to consider the technical aspect and the large economic uncertainty in the ADS.

98 TR0700316

13' International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

CALCULATION OF NEUTRON IMPORTANCE FUNCTION IN FISSIONABLE ASSEMBLIES USING MONTE CARLO METHOD

S. A. H. Feghhia'4, M. Shahriarib, H. Afarideha a Faculty of physics and nuclear sciences, AmirKabir University of Technology, Tehran, Iran bDepartment of Nuclear engineering, Shahid Beheshti University, Tehran, Iran. E-mail: [email protected].

ABSTRACT The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and super- critical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor.

Keywords: Neutron importance function; Adjoint Flux; perturbation theory; Reactor Dynamics

4 Corresponding author. Tel.: +98 21 64542591 & +98 9122109461; fax: +98 21 66495519.

99 TR0700317 I

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SENSORY SYSTEMS FOR A CONTROL ROD POSITION USING REED SWITCHES FOR THE INTEGRAL REACTOR

Je-Yong Yu, Suhn Choi, Ji-Ho Kim and Doo-Jeong Lee

Power Reactor Development Center, Korea Atomic Energy Research Institute P.O. BOX 105, Yuseong, Daejeon, 305-600, Korea Phone: 82-42-868-2835, Fax: 82-42-868-8990 E-mail: [email protected]

ABSTRACT The system-integrated modular advanced reactor (SMART) currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron free operation and the use of nuclear heating for the reactor start-up. These design features require a Control Element Drive Mechanism (CEDM) for the SMART to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is very important to the reactor safety as well as the design of the core protection system. The position indicator is classified as a Class IE component because the rod position signal of the position indicator is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and shall have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The zigzag arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. A voltage divider electrically connected to the reed switches provides a signal commensurate with the position of the control rod. The signal may then be transmitted to a position indicating device.

In order to monitor the operating condition of the rotary step motor of CEDM, the angular position detector was installed at the top of the rotary step motor by means of connecting between the planetary gear and the rotating shaft of the rotary step motor. The permanent magnet positioned in the planetary gear holder which is designed to rotate of 60° in corresponding to each step of the rotary step motor. Therefore the angular position detector can measure one step angular increase of the rotary step motor by means of detecting the angular position of the permanent magnet using the reed switch assembly outside the pressure housing of the angular position detector. The angular position detector with planetary gear can measure precisely the control rod position by considering the gear ratio and the ball screw lead as well as can check the operating condition of the rotary step motor of CEDM at real time. The two sensory systems of the position indicator and the angular position detector will cooperate to measure the position of the control rod more reliably.

100 PARALLEL SESSION 5C: ENERGY TECHNOLOGIES TR0700318

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

AMBIENT TEMPERATURE EFFECTS ON GAS TURBINE POWER PLANT: A CASE STUDY IN IRAN

Mofid Gorji and Fama Fouladi Department of Mechanical Engineering, Faculty of Engineering, University of Mazandaran, Babol, Iran E-mail: [email protected]; [email protected]

ABSTRACT Actual thermal efficiency, electric-power output, fuel-air ratio and specific fuel consumption (SFC) vary according to the ambient conditions. The amount of these variations greatly affects those parameters as well as the plant incomes.

In this paper the effect of ambient temperature as a seasonal variation on a gas power plant has been numerically studied. For this purpose, the gas turbine model and different climate seasonal variations of Ray in Iran are considered in this study. For the model, by using average monthly temperature data of the region, the different effective parameters were compared to those in standard design conditions. The results show that ambient temperature increase will decrease thermal efficiency, electric-power out put and fuel-air ratio of the gas turbine plant, whereas increases the specific fuel consumption.

Keywords: Gas turbine; Electric production; Fuel consumption

102 TR0700319

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ENERGY CONSERVATION THROUGH THE IMPLEMENTATION OF CO- GENERATION & GRID INTERCONNECTION

M. A. Dashash Manager, Power Distribution Department, Saudi Aramco E-mail: [email protected]; [email protected]

SYNOPSIS With increasing awareness of energy conservation and environmental protection, the Arab World is moving to further improve energy conversion efficiency. The equivalent of over 2.7 MM bbl is being daily burnt to fuel the thermal power plants that represent 92% of the total Arab power generation. This adds up to close to one billion barrels annually. At a conservative 30$ per barrel, this represents a daily cost of over $81 Million.

This paper will introduce two strategies with the ultimate objective to cut-off up to half of the current fuel consumption. Firstly, Cogeneration Technology is able to improve thermal efficiency from the current average of less than 25% to up to 80%. Just 1% improvement in power plant thermal efficiency represents 3 million $/day in fuel cost savings. In addition, a well-designed and operated cogeneration plant will:

• Reduce unfriendly emissions by burning less fuel as a result of higher thermal efficiency • Increase the decentralization of electrical generation • Improve the reliability of electricity supply

As an example, the Kingdom of Saudi Arabia's experience of implementing cogeneration will be presented, in particular within its hydrocarbon facilities and desalination plants. This will include the existing facilities and the planned and on-going projects.

Secondly, by interconnecting the power networks of all the adjacent Arab countries, the following benefits could be reached:

• Reduce generation reserves and enhance the system reliability • Improve the economic efficiency of the electricity power systems • Provide power exchange and strengthen the supply reliability • Adopt technological development and use the best modern technologies

At least two factors plead for this direction. On one hand, the four-hour time zone difference from Eastern to Western Arab World makes it easy to exchange power. On the other hand, this will help to reduce the reserve capacity and save on corresponding Capital investment, fuel, and O&M expenses. There are already three distinct sub-groups being interconnected in North Africa, within the Gulf countries, and in the other Middle East Arab Countries. The same benefits will be maximized by connecting the Arab Countries power pool to Europe through the MEDRING system. Reaching the mentioned objectives is not only a major contribution to energy conservation but also leads intrinsically to better environmental protection.

103 TR0700320

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

A NOVEL CO2 SEQUESTRATION SYSTEM FOR ENVIRONMENTALLY PRODUCING HYDROGEN FROM FOSSIL-FUELS

William Eucker IV United States Naval Academy, Annapolis, Maryland, USA E-mail: [email protected]

ABSTRACT Aqueous monoethanolamine (MEA) scrubbers are currently used to capture carbon dioxide (CO2) from industrial flue gases in various fossil-fuel based energy production systems. MEA is a highly volatile, corrosive, physiologically toxic, and foul-smelling chemical that requires replacement after 1000 operational hours. Room temperature ionic (RTILs), a novel class of materials with negligible vapor pressures and potentiality as benign solvents, may be the ideal replacement for MEA. Ah initio computational modeling was used to investigate the molecular interactions of ILs with CO2. The energetic and thermodynamic parameters of the RTILs as CO2 solvents are on par with MEA. As viable competitors to the present CO2 separation technology, RTILs may economize the fossil-fuel decarbonization process with the ultimate aim of realizing a green hydrogen economy.

Acknowledgement: The Office of Naval Research-Global (London) and the Chalmers University of Technology, Göteborg, Sweden, supported this research.

104 TR0700321

13lh International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

MECHANICAL ALLOYING OF MG-CO-NI POWDER FOR HYDROGEN STORAGE

Hadi Suwarno1^ Andon insani1^ Johny Wahyudi2), Eddy S. Siradj2), Bambang Herutomo1*

^Center for Nuclear Fuel Technology, National Nuclear Energy Agency, Kawasan Puspiptek, Serpong, Tangerang 15310, Banten Indonesia ^Dept. of Metallurgy and Materials, Faculty of Engineering, the University of Indonesia, Depok 16424, West Java, Indonesia E-mail:

ABSTRACT In order to develop the Mg-based materials for hydrogen storage purposes, Mg-Co-Ni alloy with the atomic ratio of the Mg: Co: Ni = 3:1:2 was prepared by mechanical alloying. The alloy was prepared from pure metal powder of magnesium, cobalt and nickel by using SPEX 8000 high energy milling (HEM) and conventional milling. Mass ratio ball to sample (B/S) were 1:1 and the milling time is varied at 5, 10, 15, 20 and 40 hours. Structure and crystallite sizes were observed by X-ray diffraction (XRD), morphology and particle size by scanning electron micrograph (SEM), and thermal properties of the sample by differential thermal analyzer (DTA).

The crystal sizes of the alloy were measured for Mg (101), Ni (200) and Co (101). Calculation results on the crystal size of the Mg exhibited that it is reduced significantly from 29 nrn into 6 nm after milling for 40 hours, while Co and Ni are slightly reduced. From the diffraction pattern of the alloy it is also showed that the peaks intensity of Mg disappears gradually, due to the amorphisation of Mg particles. It could be happened during the continuous impact between the Mg particles and the balls. A significant change of volume fraction was observed in Mg, where it changed from 62.52 % into 26.04 % after 40 hours of milling. While Co and Ni increased from 7.63 % to 10.63 % and from 25.23 % to 30.02 % respectively. The SEM results showed that the particle sizes reduce after 5 hours milling. The initial particle size of Mg was < 3.5 j^m and the final milling was reduced into 0.5 pm. In addition, agglomeration of the powder was occurred after 10 hours milling. It is due to the increase in surface area of the powder that results in the easier contact of the powders to each other. The DTA differential thermal analyses on milling time of 0 and 10 hours identified that there is an endothermic peak. The peak at 400 °C is identified as phase transition of Co from hep into fee. Weak endothermic peak encountered at temperature of 150 °C for 20 hours of milling is indicated as evaporation due to the hygroscopic properties of material. In addition to 20 hours of milling, no endothermic peak is obtained since the structure of the specimen has changed as shown in the XRD examinations.

It is concluded that milling time of 40 hours using high energy ball milling at B/S ratio of 1 can be used to produce nano size of powder before hydriding. It is predicted that nano size of powder will increase the hydriding rate compared to the original one.

Note: Hydriding system is being constructed and estimated to be finished in December 2006. Abstract will be extended after hydriding experiment is conducted.

105 TR0700322

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

CONTRIBUTION OF WIND ENERGY TO FUTURE ELECTRICITY REQUIREMENTS OF PAKISTAN

Khanji Harijana'*, Muhammad Aslam Uqailib, and Mujeebuddin Memonc aPhD Student, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro-76062, Sindh, Pakistan bProfessor, Department of Electrical Engineering, Mehran University of Engineering and Technology, Jamshoro-76062, Sindh, Pakistan cProfessor, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro-76062, Sindh, Pakistan E-mail: khanji [email protected]

ABSTRACT Pakistan is an energy deficit country. About half of the country's population has no access to electricity and per capita supply is only 520 kWh. About 67% of the conventional electricity is generated from fossil fuels with 51% and 16% share of gas and oil respectively. It has been projected that electricity demand in Pakistan would increase at an average annual growth rate of 5% to 12% under different scenarios. The indigenous reserves of oil and gas are limited and the country heavily depends on imported oil. The oil import bill is a serious strain on the country's economy and has been deteriorating the balance of payment situation. Pakistan is becoming increasingly more dependent on a few sources of supply and its energy security often hangs on the fragile threat of imported oil that is subject to supply disruptions and price volatility. The production and consumption of fossil fuels also adversely affects the quality of the environment due to indiscriminate release of toxic substances. Pakistan spends huge amount on the degradation of the environment. This shows that Pakistan must develop alternate, indigenous and environment friendly energy resources such as wind energy to meet its future electricity requirements. This paper presents an overview of wind power generation potential and assessment of its contribution to future electricity requirements of Pakistan under different policy scenarios. The country has about 1050 km long coastline. The technical potential of centralized grid connected wind power and wind home systems in the coastal area of the country has been estimated as about 484 TWh and 0.135 TWh per year respectively. The study concludes that wind power could meet about 20% to 50% of the electricity demand in Pakistan by the year 2030. The development and utilization of wind power would reduce the pressure on oil imports, protect the environment from pollution and improve the socio-economic conditions of the people.

Keywords: Wind energy; Electricity requirements; Environment friendly; Pakistan

*Corresponding author. Tel: +92-22-2653821; Fax: +92-22-2771382,

106 SESSION 6: PLENARY SESSION •- —™ -mm» mim UBJ. TR0700323

13lh International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

RECOMMENDATIONS FOR A RESTART OF MOLTEN SALT REACTOR DEVELOPMENT

Ralph W. Moir Vallecitos Research Associates, 607 E. Vallecitos Road, Livermore, CA 94550, USA E-Mail :RMoir@pacbell. net

ABSTRACT The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: • use thorium or uranium; • be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; • fission uranium isotopes and plutonium isotopes; • operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; • be a breeder or near breeder; • operate at temperature >1100 °C if carbon composites are successfully employed.

Enhancing 232U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation.

Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the number of steps to commercial deployment. Assuming electricity were worth $50 per MWe«h, then 50 years of 10 TWe power level would be worth 200 trillion dollars. If the MSR could be developed and proven for 10 B$ and would save 10% over its alternative, the total savings over 50 years would be 20 trillion dollars: a good return on investment even considering discounted future savings. The incentives for the molten salt reactor are so strong that one asks, "Why has the reactor not already been developed?"

108 TR0700324

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

INNOVATIVE ICF SCHEME - IMPACT FAST IGNITION

*M. xMurakami,1 H. Nagatomo,' T. Sakaiya,' S. Fujioka,' H. Saito,' H. Shiraga,' M. Nakai,' K. Shigemori,' H. Azechi,! M. Karasik,2 J. Gardner,2 J. Bates,2 D. Colombant,2 J. Weaver,2 A. Schmitt,2 A. Velikovich,2 Y. Aglitskiy,2 J. Sethian,2 And S. Obenshain2

1 Institute of Laser Engineering, Osaka University, 565-0871 Osaka, Japan 2Naval Research Laboratory, Washington D. C, 20375, USA E-mai 1 / mnrakami-m@ile. osaka-u. ac.jp

ABSTRACT A totally new ignition scheme for ICF, impact fast ignition (IFI), is proposed [1], in which the compressed DT main fuel is to be ignited by impact collision of another fraction of separately imploded DT fuel, which is accelerated in the hollow conical target. Two-dimensional hydrodynamic simulation results in full geometry are presented, in which some key physical parameters for the impact shell dynamics such as 108 cm/s of the implosion velocity, 200- 300 g/cmJ of the compressed density, and the converted temperature beyond 5 keV are demonstrated. As the first step toward the proof-of-principle of IFI, we have conducted preliminary experiments under the operation of GEKKO XII/HYPER laser system to achieve a hyper-velocity of the order of 108 cm/s. As a result we have observed a highest velocity, 6.5 x 107 cm/s, ever achieved. Furthermore, we have also done the first integrated experiments using the target configuration as in Fig.l and observed substantial amount of neutron yields.

Laser xL «O(nsec) Cone guide

Ablator DT layer

Laser or X-ray radiation

200 (.im

Figl. Impact fast ignition target Fig.2 X-ray streak image of the experiment with the highest velocity ever achieved.

Reference [1] M. Murakami and H. Nagatomo, Nucl. Instrum. Meth. Phys. Res. A 544 (2005) 67.

109 TR0700325

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PURSUING NUCLEAR ENERGY WITH NO NUCLEAR CONTAMINATION —FROM NEUTRON FLUX REACTOR TO DEUTERON FLUX REACTOR—

Xing Z. Li, Qing M. Wei, Bin Liu, Xie G. Zhu, Shao L. Ren Department of Physics, Tsinghua University, Beijing 100084, CHINA E-mail: [email protected]

ABSTRACT Pursuing nuclear energy with no nuclear contamination has been a long endeavor since the first fission reactor in 1942. Four major concepts have been the key issues: i.e. resonance, negative feed back, self-sustaining, nuclear radiation.

When nuclear energy was just discovered in laboratory, the key issue was to enlarge it from the micro-scale to the macro-scale. Slowing-down the neutrons was the key issue to enhance the fission cross-section in order to build-up the neutron flux through the chain-reactions using resonance between neutron and fissile materials. Once the chain-reaction was realized, the negative feed-back was the key issue to keep the neutron flux at the allowable level. The negative reaction coefficient was introduced by the thermal expansion, and the resonant absorption in cadmium or boron was used to have a self-sustaining fission reactor with neutron flux. Then the strong neutron flux became the origin of all nuclear contamination, and a heavy shielding limits the application of the nuclear energy.

The fusion approach to nuclear energy was much longer; nevertheless, it evolved with the similar issues. The resonance between deuteron and triton was resorted to enlarge the fusion cross- section in order to keep a self-sustaining hot plasma. However, the 14 MeV neutron emission became the origin of all nuclear contamination again. Deuteron plus helium-3 fusion reaction was proposed to avoid neutron emission although there are two more difficulties: the helium-3 is supposed to be carried back from the moon; and much more higher temperature plasma has to be confined while 50 years needed to realized the deuteron-triton plasma already. Even if deuteron plus helium-3 fusion plasma might be realized in a much higher temperature plasma, we still have the neutron emission from the deuteron-deuteron fusion reaction in the deuteron plus helium-3 fusion plasma. Polarized deuteron-deuteron fusion reaction was proposed early in 1980's to select the neutron-free channel in the deuteron-deuteron fusion reaction. Even if polarized deuteron has long enough life-time to keep its polarity in hot fusion plasma, there is still the probability to have the neutron emission channel from deuteron-deuteron fusion. The neutron emission in hot plasma containing deuterons is inevitable.

Isomer Hf-178 was proposed to reduce the neutron emission in terms of gamma decay controlled by X-ray. Although its reality is still in question, the resonance plays key role in this concept as well. Condensed matter nuclear science provided another chance to approach nuclear energy with no nuclear contamination. Selective resonant tunneling would select only the neutron free channel. There are five major steps in the past 17 years:

(1) Selective Resonant tunneling model has been successful to explain the 3 major puzzles in cold fusion proposed by nuclear physicist(i.e. penetration of Coulomb barrier, no neutron emission, no gamma radiation), and successful also to explain the 3 major cross-section data in hot fusion(i.e. d+t, d+d, d+He3). The Nobel prize laureate, B. Josephson of Cambridge

110 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

University, cited this theory in the famous Lindau Meeting (2004).[1,2] (2) Deuteron flux through the palladium surface at specific temperature was found correlated with heat flow in various experiments in China, Switzerland, Japan, France and Italy. [3,4] (3) The nuclear products have been confirmed in a series of experiments using deuterium flux permeating through the thin film on the palladium surface.[5] (4) Distinct from the beam-target experiment, a special procedure was proposed to search this resonance between lattice energy level and nuclear energy level. (5) Instead of the electrolytic cell, the gas loading technique has been used. It led to the discovery of the temperature of these resonances which may be as high as 1000°C. This would change greatly the usage of this nuclear energy. We may propose the future subjects of study as follows: (1) Selective Resonant tunneling model has predicted 3-deuteron fusion which has been found in experiments [6, 7] as well. The reasonable inference is the neutrino emission from the metal hydrides. The feasibility of detection of this neutrino is discussed based on the information from the KamLAND neutrino detector in Japan. (2) Nano-meter technique should be used to increase the deuterium flux through the palladium surface. (3) Gas-discharge tube in combination with optical monochrometer would be the suitable experiment at the current funding level. (4) Negative feed-back should be used to solve the problem of the reproducibility; then, based on the deuteron flux a self-sustaining reactor would be feasible.

References 1. Xing Z. Li, Jian Tian, Ming Y. Mei and Chong X. Li, "Sub-barrier Fusion and Selective Resonant Tunneling," Phys. Rev., C 61, 024610 (2000) 2. Xing Z. Li, Bin Liu, Si Chen, Qing Ming Wei, and Heinrich Hora, "Fusion cross-sections for inertial fusion energy," Laser and Particle Beam, 22,469 (2004) 3. G. Fralick , et al., "Results of an Attempt to Measure Increased Rates of the Reaction D +D— 3He + n in a Nonelectrochemical Cold Fusion Experiment," NASA Technical Memorandum 102430(1989) 4. Xing Z. Li, et al., "Correlation between abnormal deuterium flux and heat flow in a D/Pd system," J. Phys. D: Appl. Phys. 36 3095-3097, (2003) 5. Y. Iwamura, et al., "Elemental Analysis of Pd Complexes: Effects of D2 gas permeation," Jpn. J. Appl. Phys., 41 4642 (2002) 6. J. Kasagi, et al., "Energetic Protons and Alpha Particles Emitted in 150-keV Deuteron Bombardment on Deuterated Ti," J. Phys. Soc. Japan, 64 (3), 778 (1995) 7. A. Takahashi, et al., "Anomalous enhancement of three-body deuteron fusion in titanium- deuteride with low-energy D+ beam implantation," Fusion TechnoL, 34,256 (1998)

111 TR0700326

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

NUCLEAR REACTOR DEVELOPMENT IN KOREA: IT'S HISTORY AND STATUS

Jongsik Cheong, Insik Kim, Dong-Su Kim Korea Power Engineering Company, Inc. 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353, Korea E-mail: jscheong@kopec. co.kr, iskim@kopec. co.kr, dskim2@kopec. co. kr

ABSTRACT Currently in Korea, 20 nuclear plants are in operation, generating some 18,000 MWe of electricity which is about 30% of the national electricity supply. Further 8 reactors, including innovative light water reactors developed with 30 years' experience in construction and operation with continuous technology development, are either under construction or being planned. Executing an energetic program of nuclear development, Korea is now the world's sixth-ranked nuclear nation.

In this paper, at first, history of the nuclear reactor development in Korea will be discussed including technology self-reliance efforts of the nuclear industry, and future plan and prospects will also be presented. Secondly, the OPR1000 which is a Korean standard plant will be introduced in detail including its characteristics, design approach and features. Six OPRlOOO's are being operated with outstanding performance and 4 more units are under construction. The APR1400, an upgraded reactor of the OPR1000 in capacity and design, has been developed as a next generation reactor, and the contracts were signed for the first 2 units' construction in August 2006. Its development process and design features will be described. Finally, Korea's efforts for future nuclear power generation will be introduced. For future reliable energy supply, Korea has been actively participating in international cooperation such as Gen IV International Forum.

In summary, this paper will introduce the history and status of the Korean nuclear reactor development with its past, present and future, which might be helpful to understand the Korean nuclear industry and find a way for international cooperation especially with European countries.

112 PARALLEL SESSION 7A: FISSION REACTORS III TR0700327

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TRANSMUTATION OF MINOR ACTINIDES IN A CANDU THORIUM BURNER

Sümer Şahina), Şenay Yalçınb), Hacı Mehmet Şahina), Adem Acıra), Kadir Yıldızc), Necmettin Şahinc), Taner AItınokd), Mahmut Alkane)

a) Gazi Üniversitesi, Teknik Eğitim Fakültesi, Beşevler - Ankara - Türkiye E-mail: [email protected] b)Bahçeşehir Üniversitesi, Mühendislik Fakültesi, İstanbul, Türkiye c) Aksaray Üniversitesi, Mühendislik Fakültesi, Aksaray, Türkiye d)Kara Harp Okulu, Savunma Bilimleri Enstitüsü, Ankara, Türkiye e:)Niğde Üniversitesi, Mühendislik Fakültesi, Niğde, Türkiye

ABSTRACT The paper investigates the prospects of exploitation of rich world thorium reserves in CANDU reactors. Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality.

Two different fuel compositions have been selected for investigations: ® 96 % thoria (TI1O2) + 4

% PuO2 and © 91 % ThO2 + 5 % UO2 + 4 % PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the criticality k*, and the burn-up values of the reactor have been pursued by full power operation for > ~ 8 years. The reactor starts with kco = ~ 1.39 and the criticality drops down asymptotically to values koo > 1.06, still tolerable and useable in a CANDU reactor. Reactor criticality k» remains nearly constant between the 4th year and 7th year of plant operation and then a slight increase is observed thereafter, along with a continuous depletion of thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner.

Very high burn up can be achieved with the same fuel (> 160 000 MW.D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burnt up, provided that the fuel rods could be fabricated to withstand such high burn up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically. There is a great quantity of weapon grade plutonium accumulated in nuclear stockpiles. In the second phase of investigations, weapon grade plutonium is used as a booster fissile fuel material in form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure the initial criticality at startup.

Two different fuel compositions have been used: ® 97 % thoria (ThO2) + 3 % PuO2 and © 92 % 233 238 ThO2 + 5 % UO2 + 3 % PuO2. The latter is used for denaturized the new U fuel with U. The temporal variation of the criticality k*, and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k» = ~ 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the 2n year and remains above k*, > 1.06 for ~ 20 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner.

114 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

Nuclear waste actinides can also be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. In the third phase, two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: CD 95 % thoria (TI1O2) + 5 % minor actinides MAO2 and © 95 % ThO2 + 5 % MA02 + 5 % UO2. The latter allows a higher degree of nuclear safeguarding thorough denaturing the new 233U fuel with 238U. The temporal variation of the criticality k» and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k» > 1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k« = -1.06 for ~ 10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner.

Finally, in the fourth phase, a CANDU reactor fueled with a mixed fuel made of thoria (ThC>2) and the totality of nuclear waste actinides has been investigated. The mixed fuel composition has been varied in radial direction to achieve a uniform power distribution and fuel burn up in the fuel bundle.

The best fuel compositions with respect to power flattening as well as long term reactivity have been found by mixing thoria with 14 % minor actinides in form of MAO2 in the central fuel bundle and decreasing radially MAO2 content at discrete levels down to 2 % at the periphery. Furthermore, as alternative fuel, 5 % UO2 has been added to the mixed fuel for the sake of a higher degree of nuclear safeguarding through denaturing the 233U component with 238U. The temporal variation of the criticality k» and the burn-up values of the reactor have been calculated for a period of 10 years, operated at full power. The criticality starts at time zero near to k<» = ~ 1.24 for both fuel compositions. A sharp decrease of the criticality has been observed during the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2n year after sufficient 233U is accumulated and remains close to koo.end = -1.06 over - 10 years. Quasi-uniform power generation density has been realized in the fuel bundle throughout the reactor operation.

115 TR0700328 I [ 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

4+D DIGITAL ENGINEERING FOR ADVANCED NUCLEAR ENERGY SYSTEMS

Seo G. Jeong1'2, Seung K. Nam2, Kune Y. Suh1'2*

'Seoul National University, San 56-1 Sillim-dong, Gwanak-gu, 151-744, Seoul, Korea 2PHILOSOPHIA, Inc., San 56-1 Sillim-dong, Gwanak-gu, 151-744, Seoul, Korea, E-mail:[email protected]

ABSTRACT Nuclear power plants (NPPs) require massive quantity of data during the design, construction, operation, maintenance and decommissioning stages because of their special features like size, cost, radioactivity, and so forth. The system engineering thus calls for a fully automated way of managing the information flow spanning their lifecycle. In line with practice in disciplines of naval architecture, , and automotive manufacturing, the paper proposes total digital systems engineering based on three-dimensional (3D) computer-aided design (CAD) models. The signature in the proposal lies with the four-plus-dimensional (4+D) Technology™, a critical know-how for digital management. The so-called OPIUM (Optimized Plant Integrated Ubiquitous Management) features a 4+D Technology™ for nuclear energy systems engineering. The technology proposed in the 3D space and time plus cost coordinates, i.e. 4+D, is the backbone of digital engineering in the nuclear systems design and management. Based on an integrated 3D configuration management system, OPIUM consists of solutions NOTUS (Nuclear Optimization Technique Ubiquitous System), VENUS (Virtual Engineering Nuclear Ubiquitous System), INUUS (Informatics Nuclear Utilities Ubiquitous System), JANUS (Junctional Analysis Numerical Ubiquitous System) and EURUS (Electronic Unit Research Ubiquitous System). These solutions will help initial simulation capability for NPPs to supply the crucial information. NOTUS contributes to reducing the construction cost of the NPPs by optimizing the component manufacturing procedure and the plant construction process. Planning and scheduling construction projects can thus benefit greatly by integrating traditional management techniques with digital process simulation visualization. The 3D visualization of construction processes and the resulting products intrinsically afford most of the advantages realized by incorporating a purely schedule level detail based the 4+D system. Problems with equipment positioning and manpower congestion in certain areas can be visualized prior to the actual operation, thus preventing accidents and safety problems such as collision between two machines and losses in productivity. VENUS applies the virtual reality (VR) technology in nuclear industry. VR provides an interactive real-time motion with sound and tactile and other forms of feedback. Therefore, management and workers are able to comprehend the work process better by visualizing precisely how activities relate to one another, thus reducing conflicting interpretations and communication problems. VENUS can contribute to dealing with public acceptance about the NPP. Visualization of NPP can comfortably familiarize the public with existing or planned systems. INUUS provides with effective information management and methodology of expression. Information is offered variously in the field of nuclear industry for example real-time data on the spot, result of computational analysis in the process of plant design and information of documents for drawing or dividing drawing. INUUS resorts to 3D object oriented methods to efficaciously manage voluminous information with. For example, INUUS can be used as a 3D pre-processor for a NPP system analysis code. JANUS extracts the geometric data directly from the CAD files to import to multidimensional computational codes. JANUS uses these joint-CAD analysis methods so that time and efforts of the user can be minimized. EURUS combines the VR technology with a wide spectrum of high-precision, high-resolution analysis techniques and virtual design development environment so that a system can be designed and analysed in the

116 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

cyber space. EURUS realizes the cyber experiment with which the behaviour of such complex structures as NPP can be simulated and reflected onto the design. The 4+D Technology is slated to bring about revolutionary change in improving the NPPs lifecycle starting from the conceptual design to decommissioning. It will also eventually lead to paperless design and paperless plants in the near future.

117 TR0700329 i 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

FEASIBILITY STUDY OF SELF SUSTAINING CAPABILITY ON WATER COOLED THORIUM REACTORS FOR DIFFERENT POWER REACTORS

Sidik Permana*, Naoyuki Takaki, Hiroshi Sekimoto Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550, Japan Phone/ Fax: +81-3-5734-2955, E-mail: [email protected] (Sidik Permana)

ABSTRACT Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 233U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the fertile and fissile nuclides with the contribution from intermediate nuclides such as 234U and 233Pa. The conversion ratio is evaluated by considering the conversion capability of the reactor to convert the into fissile material. The fissile accumulation capability for different conditions is investigated for estimating the fissile production capability during operation. The result shows the negative reactivity coefficient, and its feasibility of breeding for different MFR and burnup. The very tight lattice pin of MFR < 0.3 is preferable for obtaining breeding condition for relatively higher burnup. The breeding capability of the reactor increases with increasing power output and decreasing power density. In relation to the self sustaining system, the large power output is easier to reach than the small power output.

118 TR0700330

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

GENERATION IV NUCLEAR PLANT DESIGN STRATEGIES

Prof. Dr. Vural Altın Bilim ve Teknik, TÜBİTAK, Ankara, Türkiye E-Mail: [email protected]

ABSTRACT In this presentation Generation IV nuclear reactor design criteria are examined under the light of known nuclear properties of fissile and fertile nuclei. Their conflicting nature is elucidated along with the resulting inevitability of a multitude of designs. The designs selected as candidates for further development are evaluated with respect to their potential to serve the different design criteria, thereby revealing their more difficult aspects of realization and the strong research challenges lying ahead.

119 TR0700331

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TYPICAL STEAM GENERATOR TUBES RUPTURE (SGTR) EFFECT ON THERMOHYDRAULIC PARAMETERS OF VVER-1000 PRIMARY LOOP

Abbas Zare, Mohammadreza Nematollahi1, Kamal Hadad, Khosro Jafarpoor, Masood Aminmozaffari Shiraz University Engineering School, Shiraz, 71345, IRAN E-mail1: [email protected]

ABSTRACT In this research variation of thermo-hydraulics parameters in primary loop under accident SGTR (Steam Generator Tube Rupture) in VVER-1000 nuclear power plant is analyzed by Rlap5/Mod 3.2 thermo-hydraulics code. In simulation of this accident, it was supposed that after establishment of steady-state condition in the system, instantaneous break with equivalent diameter of 100mm of steam generator cold collector SG#2 in the area of lower row heat exchanging tubes is created. The accident scenario is assumed as the most conservative version of damage, with loss of power supply from all off-site & in-site a.c power and a failure of two diesel generators in loop's 1, 2 of the primary cycle. The results show that after initiating of accident the pressure above the core decrease from 16 MPa to 8 MPa during 130s and the temperature in hot legs decrease from 598 k° to 528 k° of loop 2, but it decrease 503 k° in loops 1, 3 and 4. Also, primary-tosecondary coolant leakage reduces from 800 Kg/sec at the beginning of accident to lower than 50 Kg/sec after about 150 sec of the accident. The results predict correctly the behavior of main plant parameters in comparison with that are reported in PSAR shows accuracy of this model by relap5 thermohydralics code.

Key words: VVER-1000, Steam Generator, Tube Rupture, thermo-hydraulics code

120 PARALLEL SESSION 7B: HYDROGEN ENERGY TR0700332

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SOLAR PUMPED LASER AND ITS APPLICATION TO HYDROGEN PRODUCTION

K. Imasaki, T. Saiki, D. Li, S. Motokosi, *M. Nakatsuka. Institute for Laser Technology, *Institute of Laser Engineering, Osaka University. 2-6 Yamada-oka, Suita, Osaka, Japan E-mail: [email protected]

ABSTRACT Solar pumped laser has been studied. Recently, a small ceramic laser pumped by pseudo solar light shows high efficiency of more than 40% which exceeds a solar cell. Such solar pumped laser can concentrate the large area of solar energy in a focused spot of small area. This fact implies the application of such laser for clean and future renewable energy source as hydrogen.

For this purpose, 100W level laboratory solar laser HELIOS is completed using disk ceramic active mirror laser to achieve high temperature. This laser is a kind of MOP A system. Oscillator of additional small laser is used. Laser light is generated in oscillator and is amplified in ceramic disks of solar pumped. The temperature from this system is to be more than 1500K. We will use a simple graphite cavity for laser power absorption and to get a high temperature.

We are also designing a 10MW CW laser based on this technology. This may be expected an application of solar energy for hydrogen production with total efficiency of 30%.

122 TR0700333

13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NUMERICAL STUDY OF HYDROGEN ABSORPTION IN A LM-NIS HYDRIDE REACTOR

Kemal ALTINIŞIK, Muhittin TEKİN Selçuk University Engineering and Architecture Faculty, Department of Mechanical Engineering, Konya, Türkiye E-mail: kaltinisik@selcuk. edu. tr

Mahmut D. MAT Niğde University Engineering and Architecture Faculty, Deparment of Mechanical Engineering, Niğde, Türkiye

Alper ALTINIŞIK Uzel Machine Industry, Joint Stock Company, Istanbul, Türkiye

T.NejatVEZtROĞLU Clean Energy Research Institute, University of Miami Coral Gables, Florida 33124, U.S.A.

ABSTRACT Metal hydride formation in an Lm-Nis storage tank is numerically studied with a continuum mathematical model. The model considers complex heat, and mass transfer and in the reaction bed. It is found that hydride formation enchances at regions with lower equilibrium pressure. Absorbed hydrogen mass increases exponentially at earlier times of hydriding process and slow down after temperature of reaction bed increases due to the heat of reaction. Numerical results agree satisfactorily with the experimental data in the literature. Keywords: Hydride reactor, Hydrogen absorption, Numerical study

Fax:+90 332 241 06 35 Tlf:+90 332 223 19 27

123 TR0700334

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

21ST CENTURY'S ENERGY: HYDROGEN ENERGY SYSTEM

T. Nejat Veziroglu Director, Clean Energy Research Institute, University of Miami President, International Association for Hydrogen Energy E-Mail: veziroglu@unido-ichet. org

ABSTRACT Fossil fuels (i.e., petroleum, natural gas and coal), which meet most of the world's energy demand today, are being depleted fast. Also, their combustion products are causing the global problems, such as the greenhouse effect, ozone layer depletion, acid rains and pollution, which are posing great danger for our environment and eventually for the life in our planet. Many engineers and scientists agree that the solution to these global problems would be to replace the existing fossil fuel system by the Hydrogen Energy System. Hydrogen is a very efficient and clean fuel. Its combustion will produce no greenhouse gases, no ozone layer depleting chemicals, little or no acid rain ingredients and pollution. Hydrogen, produced from renewable energy (e.g., solar) sources, would result in a permanent energy system, which we would never have to change.

However, there are other energy systems proposed for the post-petroleum era, such as a synthetic fossil fuel system. In this system, synthetic gasoline and synthetic natural gas will be produced using abundant deposits of coal. In a way, this will ensure the continuation of the present fossil fuel system.

The two possible energy systems for the post-fossil fuel era (i.e., the solar hydrogen energy system and the synthetic fossil fuel system) are compared with the present fossil fuel system by taking into consideration production costs, environmental damages and utilization efficiencies. The results indicate that the solar hydrogen energy system is the best energy system to ascertain a sustainable future, and it should replace the fossil fuel system before the end of the 21st Century.

124 TR0700335

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

TECHNOLOGIES FOR HYDROGEN PRODUCTION BASED ON DIRECT CONTACT OF GASEOUS HYDROCARBONS AND EVAPORATED WATER WITH MOLTEN PB OR PB-BI

A. V. Gulevich, P. N. Martynov, V. A. Gulevsky, V. V. Ulyanov SSC RF- Institute for Physics and Power Engineering 1, Bondarenko square, Obninsk, Kaluga region, Russia, Tel: +7(48439)98014, Fax +7(48439)98072, E-mail: [email protected]

ABSTRACT Results of studies intended for the substantiation of a new energy-saving and safe technology for low cost hydrogen production have been presented. The technology's basis is direct mixing of water and (or) gaseous hydrocarbons with heavy liquid metal coolants (HLMC) Pb or Pb-Bi. Preliminary research has been done on thermal dynamics and kinetics of the processes taking place in the interaction of HLMC with hydrocarbon-containing gases. It has been shown as a result that water and gaseous hydrocarbons interact with molten Pb and Pb-Bi relatively quietly in chemical aspect (without ignition and explosions). Therefore, (and taking into account the thermal physics, physical and chemical properties of HLMC such as low pressure of saturated vapor of Pb and Pb- Bi in enhanced temperatures, their good heat conductivity and heat capacity, low viscosity, etc.) heat transfer is possible from the molten metal to water and hydrocarbons without heat transferring partitions (that is, by direct contact of the working media). Devices to implement this method of heating liquid and gaseous media provide essential advantages: - A simple design; - None heat-transferring surfaces subject to corrosion, contamination, thermal fatigue, vibration impacts; - A high effectiveness owing to a larger heat exchanging surface per volume unit; - A small hydraulic resistance. The possibility and effectiveness of heating various gaseous and liquid media in their direct contact with molten Pb and Pb-Bi has been substantiated convincingly by experimental results at IPPE. Besides, ttie following processes of hydrogen-containing media conversion have been proved feasible thereby. 1. Water decomposition into hydrogen and oxygen. The process can develop at temperatures of 400-1000°C. It is necessary to provide constant removal of oxygen from the reaction zone and maintain a minimum possible content of chemically active oxygen in the melt. 2. Pyrolytic decomposition of hydrocarbons into carbon and hydrogen (at t>500°C). A valuable product is formed in this process - powdery carbon readily removable from the reaction zone owing to a large density difference of carbon vs. liquid metal. 3. The oxidation conversion of hydrocarbons (at t>500°C). Hydrogen and CO2, hydrogen and synthetic gas (H2 and CO mixture) can be obtained as end products. This process develops more effectively compared to the traditional vapor conversion. The increase of conversion effectiveness is caused by the new processes not employed before: hydrocarbon oxidation by oxides present in the reaction zone as dissolved in the melt and in solid phase; co-oxidation of hydrocarbons by evaporated water and HLMC oxides.

125 13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

As a result of enhanced effectiveness of oxidation conversion, the conditions for its fulfillment can be considerably simplified - the working pressure, as well as the process temperature can be decreased to the level at which it becomes possible to use structural materials, equipment, and appropriate measures for the employment of technology with coolants that have been developed and substantiated for operation in circuits with Pb and Pb-Bi coolants. The dimensions of «direct- contact» devices for hydrogen production can be very small. Therefore, they can find application both in large-scale hydrogen production, and in small-size (remote) sources of hydrogen.

126 TR0700336

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

THERMAL DESIGN AND TECHNICAL ECONOMICAL AND ENVIRONMENTAL ANALYSES OF A HYDROGEN FIRED MULTI-OBJECTIVE COGENERATION SYSTEM

Ali DURMAZ*, M. Zeki YILMAZOGLU*, Ayşegül PASAOGLU* *Gazi Üniversitesi, Mühendislik-Mimarhk Fak., Makine Mühendisliği Böl. Ankara Türkiye E-mail: [email protected], [email protected], [email protected]

ABSTRACT Approximately 85% of rapidly increasing world energy demand is supplied by fosil fuels. Extreme usage of fossil fuels causes serious global warming and environmental problems in form of air, soil and water pollutions. The period, in which fossil fuel reserves are decreasing, energy costs are increasing rapidly and new energy sources and technologies do not exist on the horizon, can be called as the expensive and critical energy period. Hydrogen becomes a matter of primary importance as a candidate energy source and carrier in the critical energy period and beyond to solve the energy and environmental problems radically. In this respect, the main obstacle for the use of hydrogen is the high cost of hydrogen production, which is expected to be decreased in the feature.

The aim of this study is to examine how hydrogen energy will be able to be integrated with the existing energy substructure with technical and economical dimensions. In this sense, a multi objective hydrogen fired gas turbine cogeneration system is designed and optimized. Technical and economical analyses depending on the load conditions and different hydrogen production cost are carried out. It is possible that the co-generated heat is to be marketed for residence and industrial plants in the surrounding at or under market prices. The produced electricity however can only be sold to the public grid at a high unit support price which is only obtainable in case of the development of new energy technologies. This price should however be kept within the nowadays supportable energy price range. The main mechanism to be used during the design stage of the system to achieve this goal is to decrease the amortization and operational costs which lead to decrease investment and fuel costs and to increase the system load factor and co- generated heat revenues.

127

PARALLEL SESSION 7C FUSION & PHYSICS TR0700337

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

RADIATION-PHYSICAL PROCESSES AT NUCLEAR-TRANSMUTATION OF SILICON DOPED BY PALLADIUM

Sh. Makhkamov, N. A.Tursunov, A. R. Sattiev Institute of Nuclear Physics of Uzbek Academy of Sciences, Tashkent, Uzbekistan

ABSTRACT It is known that at the nuclear-transmutation of doped silicon, it takes place not only transformation of 30Si into 31P, but also doped impurities to corresponding isotopes. The latter may significantly influence to properties of the nuclear-transmutated material by changing their initial positions in the silicon lattice and by following migration in the material, which may be also stimulated by radiation. This work devoted for studying peculiarities of physical processes that occurs in Si at nuclear transmutation, identification and revealing microstructure of defects, degree of homogeneity of their distribution, as well as interactions between nuclear- transmutated isotopes with Pd and its influence on properties of the material.

Single crystal n- and p-type silicon crystals with resistivity from 1 to 40 Ohm-cm, dislocation density of ~104 cm"2 and oxygen content of ~1017 cm"3 were used in our studies. Doping of silicon wafers by Pd was performed by a thermal diffusion technique in the temperature range of 1050-^1250°C during 0.5-^5 hrs. Irradiation was conducted by nuclear reactor neutrons with fluences of 5-10184-5-1019 cm"2 with the followed annealing at 800°C during 30 min. Electrical, spectrometric and X-ray fluorescent analysis methods were used to reveal efficiency of formation of impurity centers before and after nuclear transmutation, types and states of impurity-defect centers and their influence upon electrophysical, photoelectrical and recombination parameters of doped silicon. On the base of the conducted research it was obtained that after Pd diffusion Pd- related acceptor (Ec - 0.18 and Ev + 0.34 eV) levels and donor level Ev + 0.32 eV are formed in the silicon forbidden gap.The neutron irradiation of doped silicon causes nuclear transformation of isotopes 102Pd and Pd to 103Pd with the followed electron capture, leading to formation of stable isotope 103MRh. It was shown that the irradiation of Si by neutron fluences of 5-1018 •*• 19 2 5-10 cm" and followed annealing leads to formation of Rh-related levels Ec-0.32 and Ec-0.53 3B, along with those of Pd, concentrations of which increases with increase of the neutron fluences. In this work we also discuss possible mechanism of interactions between Pd, Rh and radiation defects at neutron-transmutation conditions in comparison with that in diffusional doping. This work was supported by the grant O-2.1.23 of the Center for Science and Technology at Cabinet of Ministry of Republic of Uzbekistan and the grant NelO-06 of Fund of Uzbekistan Academy of Sciences for Support Fundamental Research.

130 _. ._„ uu IUBa UMI H TR0700338

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

LASER-DRIVEN HIGH-ENERGY IONS AND THEIR APPLICATION TO INERTIAL CONFINEMENT FUSION M. Borghesi School of Mathematics and Physics, The Queen's University of Belfast (UK) E-mail: [email protected]

ABSTRACT The acceleration of high-energy ion beams (up to several tens of MeV per nucleon) following the interaction of short and intense laser pulses with solid targets has been one of the most important results of recent laser-plasma research [1]. The acceleration is driven by relativistic electrons, which acquire energy directly from the laser pulse and set up extremely large (~TV/m) space charge fields at the target interfaces. The properties of laser-driven ion beams (high brightness and laminarity, high-energy cut-off, ultrashort burst duration) distinguish them from lower energy ions accelerated in earlier experiments at moderate laser intensities, and compare favourably with those of "conventional" accelerator beams.

In view of these properties, laser-driven ion beams can be employed in a number of innovative applications in the scientific, technological and medical areas. We will discuss in particular aspects of interest to their application in an Inertial Confinement Fusion context. Laser-driven protons are indeed being considered as a possible trigger for Fast Ignition of a pre- compressed fuel. [2] Recent results relating to the optimization of beam energy and focusing will be presented. These include the use of laser-driven impulsive fields for proton beam collimation and focusing [3], and the investigation of acceleration in presence of finite-scale plasma gradient. Proposed target developments enabling proton production at high repetition rate will also be discussed.

Another important area of application of proton beams is diagnostic use in a particle probing arrangement for detection of density non-homogeneities [4] and electric/magnetic fields [5]. We will discuss the use of laser-driven proton beams for the diagnosis of magnetic and electric fields in planar and hohlraum targets and for the detection of fields associated to relativistic electron propagation through dense matter, an issue of high relevance for electron driven Fast Ignition.

[1] M.Borghesi, et al., Fusion Science and Technology, 49, 412 (2006) [2] M.Roth, M.Borghesi et al, Phys. Rev. Lett., 86,436 (2001) [3] T. Toncian, et a/._Science, 312, 410 (2006) [4] AJ.Mackinnon, P.K.Patel, M.Borghesi et al, Phys Rev. Lett., 97, 045001 (2006) [5] M.Borghesi et al, Phys. Plasmas, 9, 2214 (2002)

131 TR0700339 i j_ 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PLASMA FOCUS SYSTEM; DESIGN, CONSTRUCTION AND EXPERIMENTS

A. Alaçakır3, Y. Akgüna, A. S. Bölükdemir3, A. Elmahb, H. Karadeniz2, T. Öncüa, E. Recepoğlua, İ. T. Çakır8, Ö. Yeşiltaş8, S. Zararsız3 a) Sarayköy Nuclear Research and Training Center (SANAEM) istanbul Road 30. km. 06983 Saray- Kazan/ ANKARA b) Çekmece Nuclear Research and Training Center, Altmşehir Road, 5. km. Halkah/İSTANBUL E-mail:

ABSTRACT The aim of this work is to construct a compact experimental system for fusion research. The design, construction and experiments of the 3 kJ Mather type plasma focus machine is described. This machine is established for neutron yield and fast neutron radiography by D-D reaction which is given by D+ D->iHe (0.82 MeV) + n (2.45 MeV). Investigation of the geometry of plasma focus machine in the presence of high voltage drive and vacum system setup is shown. 108 neutron per pulse and 200 kA peak current is obtained for many shots. Scintillator screen for fast neutron imaging, sensitive to 2.45 MeV neutrons, is also manufactured in our labs. Structural neutron shielding computations for safety is also completed.

Keywords: Plasma focus, neutron, D-D reaction, fusion research

132 TR0700340

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NEW DISCOVERY: QUANTIZATION OF ATOMIC AND NUCLEAR REST MASS DIFFERENCES

Fangil A. Gareev and Irina E. Zhidkova

Joint Institute for Nuclear Research, Dubna, Russia E-mail: gareev@thsun 1 .jinr. ru

ABSTRACT We come to the conclusion that all atomic models based on either the Newton equation and the Kepler laws, or the Maxwell equations, or the Schrodinger and Dirac equations are in reasonable agreement with experimental data. We can only suspect that these equations are grounded on the same fundamental principle(s) which is (are) not known or these equations can be transformed into each other. We proposed a new mechanism of LENR: cooperative processes in the whole system - nuclei + atoms + condensed matter - nuclear reactions in plasma - can occur at smaller threshold energies than the corresponding ones on free constituents. We were able to quantize [1] phenomenologically the first time the differences between atomic and nuclear rest masses by the formula: AAM = -^--0.0076294 (in MeV/c2), n, =1,2,3,... Note that this quantization rule is justified for atoms and nuclei with different A, N and Z and the nuclei and atoms represent a coherent synchronized open systems - a complex of coupled oscillators (resonators). The cooperative resonance synchronization mechanisms are responsible for explanation of how the electron volt world can influence on the nuclear mega electron volt world. It means that we created new possibilities for inducing and controlling nuclear reactions by atomic processes grounded on the fundamental low of physics - conservation law of energy.

The results of these research fields can provide new ecologically pure mobile sources of energy independent from oil, gas and coal, new substances, and technologies. For example, this discovery gives us a simple and cheep method for utilization of nuclear waste.

References [ 1 ] F.A. Gareev, I.E. Zhidkova, E-print arXiv Nucl-th/0610002 2006.

133 TR0700341

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

THE ALBEDO PROBLEM FOR PURE-QUADRATIC SCATTERING

D. Tiireci1 and R. G. Tiireci2 'Ankara University, Faculty of Science, Dept. of. Physics, 06100, Ankara Türkiye 2Kırıkkale University, Faculty of Arts and Science, Dept. of Physics, 71450, Kırıkkale, Türkiye

E-mail2: dtureci@science. ankara. edu. tr

ABSTRACT The one speed and time-independent neutron transport equation can be considered for a homogeneous medium which thickness is identified as t in plane geometry and has nucleus in it. Here, interaction of nucleuses which are in slab that its thickness is t and neutrons that are incoming to the medium from outside is considered as depend on quadratic quantities. In other words, neutron-nucleus interactions are proportional with second degree of neutrons advent direction(ju) and scattering direction(//). The Case's eigenfunctions and the orthogonality relations of them can be written for this scattering. In addition the half-space albedo problem can be investigated with the same way.

In this study, the singular eigenfunctions method is used. For the predicted neutron fluxes over surfaces the albedo and the transmitting relations can be written easily. Thus the albedo and the transmitting values for the slab and the albedo values for the half-space can be found as numerical. To examine the accuracy of the results calculated data can be worked on by interpolation method. Thus, while the quadratic anisotropic coefficient goes to zero, the results are to be reduced to results in isotropic scattering condition. The interpolated results, which are calculated in this way, are achieved so convergent to isotropic results.

134 PARALLEL SESSION 8A: FISSION REACTORS IV TR0700342

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

NOBLE GAS, BINARY MIXTURES FOR COMMERCIAL GAS-COOLED REACTOR SYSTEMS

Mohamed S. El-Genk and Jean-Michel Tournier Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept. The University of New Mexico, Albuquerque, NM 8713, USA (505) 277 - 5442, E-mail: [email protected]

ABSTRACT Commercial gas cooled reactors employ helium as a coolant and working fluid for the Closed Brayton Cycle (CBC) turbo-machines. Helium has the highest thermal conductivity and lowest dynamic viscosity of all noble gases. This paper compares the relative performance of pure helium to binary mixtures of helium and other noble gases of higher molecular weights. The comparison is for the same molecular flow rate, and same operating temperatures and geometry. Results show that although helium is a good working fluid because of its high heat transfer coefficient and significantly lower pumping requirement, a binary gas mixture of He-Xe with M = 15 gm/mole has a heat transfer coefficient that is ~7% higher than that of helium and requires only 25% of the number stages of the turbo-machines. The binary mixture, however, requires 3.5 times the pumping requirement with helium. The second best working fluid is He-Kr binary mixture with M= 10 gm/mole. It has 4% higher heat transfer coefficient than He and requires 30% of the number of stages in the turbo-machines, but requires twice the pumping power.

136 TR0700343

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ACR-1000: OPERATOR-BASED DEVELOPMENT

B.Shalaby and Ali Alizadeh

Atomic Energy of Canada Limited 2251 Speakman Drive, Mississauga, Ontario, Canada L5K 1B2 E-mail: [email protected]

ABSTRACT Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU®* reactors to establish Generation III+ Advanced CANDU Reactor ™ (ACR™) technology. The ACR-1000™ nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability.

The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution.

Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDU™ technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants.

As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of main program elements, engineering, R&D, and demonstration and experience feedback. * CANDU® is a registered trademark of Atomic Energy of Canada Limited (AECL).

** Advanced CANDU Reactor™, ACR™ , ACR-1000™ and SMART CANDU™ are trademarks of

AECL.

137 TR0700344

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PLUTONIUM AND MINOR ACTINIDES MANAGEMENT IN THERMAL HIGH- TEMPERATURE REACTORS - THE EU FP6 PROJECT PUMA

James C. Kuijper5 et al.6 5RG, Westerduinweg 3, NL-1755 ZG Petten, The Netherlands E-mail: [email protected]

ABSTRACT The High Temperature gas-cooled Reactor (HTR) can fulfil a very useful niche for the purposes of Pu and Minor Actinide (MA) incineration due to its unique and unsurpassed safety features, as well as to the attractive incentives offered by the nature of the coated particle (CP) fuel. No European reactor of this type is currently available, but there has been, and still is, considerable interest internationally. Decisions to construct such a reactor in China and in South Africa have already been made or are about to be made. Apart from the unique and unsurpassed safety features offered by this reactor type, the nature of the CP fuel offers a number of attractive characteristics. In particular, it can withstand burn-ups far beyond that in either LWR or FR systems. Demonstrations as high as 75% FIMA have been achieved. The coated particle itself offers significantly improved proliferation resistance, and finally with a correct choice of the kernel composition, it can be a very effective support for direct geological disposal of the fuel.

The overall objective of the PUMA project, a Specific Targeted Research Project (STREP) within the European Union 6th Framework (EU FP6), is to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO2-free energy generation.

A number of important issues concerning the use of Pu and MA in gas-cooled reactors have already been dealt with in other projects, or are being treated in ongoing projects, e.g. as part of EU FP6. However, further steps are required to demonstrate the potential of HTRs as Pu/MA transmuters based on realistic/feasible designs of CP Pu/MA fuel and the PUMA focuses on necessary key elements, which are not covered by these other projects. Earlier projects indicate favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs will be investigated to optimise the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprises the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR.

It is also envisaged to optimise present Pu CP design and to explore feasibility for MA fuel. New CP designs will be explored that can withstand very high burn-ups and are well adapted for disposal after irradiation. The project benefits greatly from access to past knowledge from

5 Corresponding author, NRG, Westerduinweg 3, NL-1755 ZG Petten, The Netherlands, [email protected] 6 Further authors from the PUMA partner organisations; to be specified further in the final paper

138 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Belgonucleaire's Pu HTR fuel irradiation tests of the 1970-s, and also secures access to materials made at that time.

(Very) High Temperature Reactor (VHTR) Pu/MA transmuters are envisaged to operate in a global system of various reactor systems and fuel cycle facilities. Fuel cycle studies are envisaged to study the symbiosis to LWR, GCFR and ADS, and to quantify waste streams and radiotoxic inventories. The technical, economic, environmental and socio-political impact will be assessed as well. The PUMA project runs from September 1, 2006, until August 31, 2009, and is being executed by a consortium of 15 European partner organisations and one from the USA. The paper presents an overview of planned activities and preliminary/expected results.

139 TR0700345

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

DESIGN STUDY ON SMALL CANDLE REACTOR

Hiroshi Sekimoto and Mingyu Yan Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology E-mail:

ABSTRACT A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required. This burnup strategy can derive many merits. The change of excess reactivity along burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Therefore, the operation of the reactor becomes much easier than the conventional reactors especially for high burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. In our previous works it is appeared that application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burnup of the spent fuel is about 40 % that is equivalent to 40 % utilization of the natural uranium without the reprocessing and enrichment. This reactor can be realized for large reactor, since the neutron leakage becomes small and its neutron economy becomes improved. In the present paper we try to design small CANDLE reactor whose performance is similar to the large reactor by increasing its fuel volume ration of the core, since its performance is strongly required for local area usage. Small long life reactor is required for some local areas. Such a characteristic that only natural uranium is required after second core is also strong merit for this case. The core with 1.0m radius, 2.0m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From equilibrium analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution etc. The burn-up velocity is less than l.Ocm/year that enables a long life design easily. The core averaged discharged fuel burn up depth is about 40%. For more understanding about the effect of the coolant to fuel volume ratio, the comparison between five cases is made. The coolant channel radius is different from each other, while fuel pin pitch is fixed. Further, the comparison is made with fixed coolant channel radius and different fuel pin pitches.

140 TR0700346

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

MONTE CARLO BENCHMARK CALCULATIONS FOR 400MWTH PBMR CORE

Hong-Chul KIM1, Soon Young KIM2, Jong Kyung KIM1'*, and Jae Man NOH3 Department of Nuclear Engineering, Hanyang University, 17 Haengdang, Seongdong, Seoul, 133-791, Korea 2 Innovative Technology Center for Radiation Safety, Hanyang Univ., 17 Haengdang, Seongdong, Seoul, 133-791,Korea 3 Korea Atomic Energy Research Institute, 150 Deockjin, Yuseong, Daejeon, 305-353, Korea x* Tel) +82-2-2220-0464, Fax) +82-2-2294-4800, E-mail: [email protected]

ABSTRACT A large interest in high-temperature gas-cooled reactors (HTGR) has been initiated in connection with hydrogen production in recent years. In this study, as a part of work for establishing Monte Carlo computation system for HTGR core analysis, some benchmark calculations for pebble-type

HTGR were carried out using MCNP5 code. The core of the 400MWth Pebble-bed Modular Reactor (PBMR) was selected as a benchmark model. Recently, the IAEA CRP5 neutronics and thermal-hydraulics benchmark problem was proposed for the testing of existing methods for HTGRs to analyze the neutronics and thermal-hydraulic behavior for the design and safety evaluations of the PBMR. This study deals with the neutronic benchmark problems, for fresh fuel and cold conditions (Case F-l), and first core loading with given number densities (Case F-2), proposed for PBMR. After the detailed MCNP modeling of the whole facility, benchmark calculations were performed. Spherical fuel region of a fuel pebble is divided into cubic lattice element in order to model a fuel pebble which contains, on average, 15000 CFPs (Coated Fuel Particles). Each element contains one CFP at its center. In this study, the side length of each cubic lattice element to have the same amount of fuel was calculated to be 0.1635cm. The remaining volume of each lattice element was filled with graphite. All of different 5 concentric shells of CFP were modeled. The PBMR annular core consists of approximately 452000 pebbles in the benchmark problems. In Case F-l where the core was filled with only fresh fuel pebble, a BCC(body-centered-cubic) lattice model was employed in order to achieve the random packing core with the packing fraction of 0.61. The BCC lattice was also employed with the size of the moderator pebble increased in a manner that reproduces the specified F/M ratio of 1:2 while preserving the packing fraction of 0.61 in Case F-2. The calculations were pursued with ENDF/B-VI cross-section library and used sab2002 S(a, P) thermal cross-section library for

graphite material. The resulting keff was calculated to be 1.27949±0.00052 and 1.14014±0.00055 for Case F-l and F-2, respectively. Comparing with other previous results from MCNP4b code, these results gave an agreement of keff difference by 141 pcm and 1140 pcm for Case F-l and F- 2, respectively. There results were caused by a different geometry modeled. While 3 bottom cone regions and de-fueling chutes were modeled explicitly in this study, these were assumed to the surfaces flattened in MCNP4b calculations. This study can be contributed and utilized directly in the establishment of benchmark problems to develop deterministic neutronics analysis tools and methods, which lagged behind the state of the art compared to other reactor technologies, to design and analyze PBMR. It is also expected that this study would be utilized in the validation of a deterministic computer code for HTGR core analysis which will be developed in near future in Korea.

141

PARALLEL SESSION 8B: LASERS & NUCLEAR REACTORS TR0700347

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PULSE REACTOR SYSTEM FOR NUCLEAR PUMPED LASER USING LOW ENRICHED URANIUM

Toru Obara*, Hiroki Takezawa Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-19, Ookayama, Meguro-ku, Tokyo, 152-8550, Japan *) Corresponding author: Phone: +81-3-5734-2380, Fax: +81-3-5734-2959, E-mail: [email protected]

ABSTRACT Nuclear Pumped Laser (NPL) is a laser technology in which laser oscillation is done by energy of nuclear reaction. It is expected that NPL by nuclear fission make it possible to broaden the use of nuclear energy, which is now limited to electricity generation and the use of nuclear heat. It is often call as Reactor Pumped Laser (RPL), but in this paper, it is called as NPL simply. In NPL (RPL), fission fragments by nuclear fissions fly through Ar-Xe gas medium in laser cells, and the kinetic energy of the fission fragments do pumping of the laser oscillation. One of the designs of nuclear reactor system for NPL is coupling of high enriched uranium metal pulse cores and laser oscillation cells, whose inside surface is coated with high enriched uranium. Laser oscillation experiments have been performed using this type of system in IPPE, Russia, and it has been shown the laser oscillation was possible. The use of high enriched uranium is limited in the view point of non-proliferation. For research and even for practical use of NPL, reactor system using low enriched uranium is needed.

In the paper, the possibility of pulse reactor system for NPL using low enriched uranium is discussed. For the neutronic analysis, continuous energy Mote Carlo code MVP was used with JENDL-3.3 nuclear data library. By the analysis, it became clear that it was possible to design a reactor system with 20% enriched uranium by making the size of pulse cores large and by making the position and number of the cores proper. The system is possible to give enough energy to the laser oscillation cells for laser pumping. It is also shown that this system has more flatten power distribution in laser oscillation cells, which is desirable for laser pumping. For the detail analysis of laser pumping, kinetic analysis in pulse operation is needed. For the analysis, time dependent neutron coupling factors were needed between each reactor region. For the calculation, modified MVP Mote Carlo code was developed. The modified code can estimate the factors by the time between neutron fissions in different regions. The method and the results are discussed in the paper also. By the study, it becomes clear that it is possible to give enough energy to laser oscillation cells using 20% enriched uranium for pulse cores and for coating inside the laser cells. The use of low enriched uranium will solve the problem of limitation in research and practical use of NPL.

144 TR0700348

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

DESIGN CHARACTERISTICS OF REGIONAL ENERGY REACTOR, REX-10

Jong-Won Kim, Hyoung-Kyoun Ahn (Seoul National Univ.), Hyeong-Min Joo, Byeong-IH Jang (Hanyang Univ.), Myung-Hyun Kim (Kyunghee Univ.), Han-Gyu Joo, Moo-Hwan Kim and Goon-Cherl Park (Seoul National Univ.) E-mail: [email protected]

ABSTRACT Regional Energy Research Institute for the Next Generation (RERI) is to develop a small-scale electric power system with an integrated transmission/distribution/load powered by an environmentally-friendly and stable small nuclear reactor. The newly designed REX-10 (Regional Energy Reactor, lOMWth) has been developed to maintain system safety in order to be placed in densely populated region, island, etc. The design objectives are inherent safety, non- proliferation and economical Efficiency. For high safety, natural circulation, pool-type vessel and low operation pressure (2.0 MPa) are introduced. In addition, the thorium fuel cycle with 10 year lifetime without exchanging fuel is considered for the sake of the non-proliferation. Moreover, the economical efficiency is ensured by the unmanned automatic control. The system pressure and capacity are determined properly for regional energy reactor. The operation pressure is 2.0 MPa and the thermal power is 10 MWth. The major research activities for REX-10 design are natural circulation, steam-gas pressurizer and thorium fuel cycle. From the viewpoint of design characteristic for regional energy reactor, inherent safety and passive system should be introduced. Thus, the natural circulation system and self-pressurized operation by steam-gas pressurizer are adopted. In this paper, REX-10 and its design objectives and characteristics are introduced. Moreover, the detail studies of the major research activities for REX-10 design are presented.

145 TR0700349

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

APPLICATIONS OF SUPER-HIGH INTENSITY LASERS IN NUCLEAR ENGINEERING

Rainer Salomaa, Antti Hakola, and Marko Santala Advanced Energy Systems Helsinki University of Technology POBox 4100, FIN-02015 TKK, Finland E-mail: [email protected]

ABSTRACT Laser-plasma interactions arising when a superintense ultrashort laser pulse impinges a solid target creates intense partly collimated and energy resolved photons, high energy electron and protons and neutrons. In addition the plasma plume can generate huge magnetic and electric fields. Also ultra short X-ray pulses are created. We have participated in some of such experiments at Rutherford and Max-Planck Institute and assessed the applications of such kind as laser-driven accelerators. This paper discusses applications in nuclear engineering (neutron sources, , fast ignition and transmutation, etc). In particular the potential for extreme time resolution and to partial energy resolution are assessed.

146 TR0700350

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

A STUDY ON PHYSICAL CHARACTERISTICS OF SUPERCRITICAL LIGHT- WATER REACTOR LOADED WITH (233U-TH-238U) OXIDE FUEL

E.G.Kulikov1, A.N.Shmelev1, G.G.Kulikov2, V.A.Apse1 'Moscow Engineering Physics Institute (State University), Russia E-mail: [email protected] international Science and Technology Center, Russia, E-mail: [email protected]

ABSTRACT The attractiveness of using (U-Th)-fuel in supercritical light water reactor is considered. The dilution of2 3U in 238U is proposed with the purpose of increasing non-proliferation of this fissile isotope. Comparison of different fuel compositions is accomplished from the point of view of fissile isotope breeding and achieved burn-up; parasitic neutron absorption cross-sections are also compared. It is analyzed the impact for neutron balance of both cladding materials: zirconium alloy and stainless steel.

147 TR0700351 | i 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

NEUTRONICS OF A LIQUID SALT COOLED - VERY HIGH TEMPERATURE REACTOR

J. Zakova Nuclear and Reactor Physics, KTH, Sweden, E-mail: [email protected]

ABSTRACT Background During last few years, the interest in the innovative, Liquid Salt cooled - Very High Temperature Reactor (LS-VHTR), has been growing. The preconceptual design of the LS-VHTR was suggested in Oak Ridge National Laboratory (ORNL) [1] and nowadays, several research institutions contribute to the development of this concept. The LS-VHTR design utilises a prismatic, High Temperature Reactor (HTR) fuel [2] in combination with liquid salt as a coolant. This connection of high-performance fuel and a coolant with enhanced heat transfer abilities enables efficient and economical operation. Main objective of the LS-VHTR operation may be either an efficient electricity production or a heat supply for a production of hydrogen or, combination of both. The LS-VHTR is moderated by graphite. The graphite matrix of the fuel blocks, as well as the inner and outer core reflectors serve as a thermal buffer in case of an accident, and they provide a strong thermal feedback during normal reactor operation. The high inherent safety of the LS-VHTR meets the strict requirements on future reactor systems, as defined by the Gen IV project.

This work, purpose, scope, contribution to the state-of-art The design, used in the present work is based on the first ORNL suggestion [1]. Recent study is focused on comparison of the neutronic performance of two types of fuel in the LS-VHTR core, whereas, in all previous works, only uranium fuel has been investigated. The first type of fuel, which has been employed in the present analysis, is based on the spent Light Water Reactor (LWR) fuel, whereas the second one consists of enriched uranium oxide. The results of such a comparison bring a valuable knowledge about limits and possibilities of the LS-VHTR concept, when employed as a spent fuel burner.

Method Figure 1 shows a 3-D drawing of the LS-VHTR core, which contains 324x10 hexagonal fuel blocks. Each fuel block contains 216x10 fuel pins, which consists of TRISO particles incorporated into a graphite matrix. The external radius of a TRISO particle has been set to 410um, the radius of the fuel kernel to 150 urn, in case of plutonium fueled core, and 215 um in case of uranium fueled core. The core was modelled in stochastic, three-dimensional code MCNP, version 4c3, in the finest detail. First, an undermoderated core setup was found for both types of fuel by modifying the fuel to moderator ratio; then, the void and the thermal coefficients of reactivity were investigated. Few single - component molten salts were involved in the study of the void effect, in order to estimate worth of these components; NaF, BeF2, LiF, ZrF4. As a reference multi-component salt, Lİ2Be4F, referred to as FLiBe, was investigated.

148 13th Internationa] Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Results Table 1 summarizes the total void worth achieved with the individual liquid salt components and

FLiBe. It can be seen, that removing BeF2 from the core brings a negative reactivity contribution, while other three components, NaF, LiF and ZrF4 would in a mixture contribute to the reactivity positively. Voiding FLiBe, which is a mixture of 66% of LiF and 34% BeFi, is equivalent to a negative reactivity insertion. Both the moderator and the fuel temperature coefficients of reactivity are large and negative for both plutonium and uranium fueled core. In the operational temperature interval (1200 K for graphite and 1500 K for fuel), the total temperature feedback is - 7.82 pcm/K for the plutonium fueled core and -2.47 pcm/K for the uranium fueled core. This results show, that the LS-VHTR core has a potential to meet the basic safety requirements as both uranium, and spent LWR fuel burner.

Figures and Tables

t Riser I

Axial Rdleetor

Radial keikelor: solid graphite

Kiwi Blocks k Fuel Pail: 7 >.! cm horizontally: 324 blocks vertically: 10 blocks

ial Reflector, i mm ('oolaiit Space graphite hioeks around each block CwJaul Riser II

Figure 1: Core of LS-VHTR

Table 1 : Total void worths for selected liquid salts and core configurations Plutonium Fuel, kernel 150 um, pf Uranium Fuel; kernel 215 fim, pf Liquid 14.45% 24.32%, enrichment 15% Salt m [pem] 1 ($j [pem]

Li2BeF4 -2.40 -641 -0.42 -308

BeF4 -5.26 -1405 -1.61 -1194

Lİ2 +0.31 +83 +0.61 +449 NaF +9.48 +2530 +7.50 +5545

ZrF4 +4.61 + 1230 +2.09 + 1551

149 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

References

[1] D. T. Ingersoll, L. J. Ott, J. P. Renier, S. J. Ball, W. R. Corwin, C. W. Forsberg, D. F. Williams, D. F. Wilson, L. Reid, G. D. Del Cul, P. F. Peterson, H. Zhao, P. S. Pickard, E. J. Parma. Status ofPreconceptual Design of the Advanced High-Temperature Reactor. (ORNL, The United States of America, Tennessee 2004).

[2] A. Talamo, W. Gudowski, F. Venneri, Annals of Nuclear Energy, 31, 173-196 (2004), The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB.

150 PARALLEL SESSION 8C: NUCLEAR REACTION EFFECTS TR0700352

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

THE POSSIBILITY OF INTRODUCTION THE TEMPORAL SCALE HUBBLE IN DYNAMICAL REACTION AU+AU AT 200 AGEV STUDIED IN BRAHMS EXPERIMENT (BROOKHAVEN - USA)

C. Besliu1, A. Danu2, Al. Jipa1,1. S. Zgura2 'University of Bucharest, Faculty of Physics, Romania institute of Space Science, Bucharest, Romania E-mail: [email protected]

ABSTRACT The main idea is the introduction at scale (fm/c~10~23s) of the nuclear processes of temporal relation Hubble. It is obviously that an experimental timer for nuclear reactions having the unit ~10"23s is very difficult to be introduced. The nuclear physics at low energies established only for same particularly cases the possible evaluation of time: the "Doppler shift method" (10~12-10~14s) and the "chambering effect" (10~18-10~23s). In the field of very high energy the temporal scale is crucial for enlarging and explaining the meaning of phase transitions of hadronic matter, the moments of quasi thermal and thermal equilibrium for QGP and its period, start for the production of different resonances and clusters.

We can add too the theoretical need for a temporal scale: the solutions experimentally valuable for the hydrodynamic state equation can be calculated only when the hot plasma evolves after the critical time of equilibrium. Tell now some estimation for the nuclear time was done in the particular cases using the geometrical model formulated for nuclear collisions at high energies [] (the time production of protons, kaons and pions). The main ask (problem) putted by the experiment is the construction of a physical time scale having as "start" the creation of the initial fireball and quark-gluon plasma and as stop all the interactions marked by the "take off of participants (the "freeze-out" moment). The temporal evolution could allow extrapolating our physical information dominated by the freeze-out state in the past period near the initial explosion. To elaborate this scenario our group was inspired by the strong connections between the high energy physics and the cosmology and astrophysics. The implications of the neutrino physics inside the general relativity changing drastically the scenario remains, the obligatory presence of the quark-gluon plasma immediately after the big bang, the similitude "big-bang" - fireball, represent the strong argument for our attempting.

Because the simplicity which requires pertinent answers we choose the cosmologic algebra created by Einstein, De Glitter, Friedman for the space-time evolution of universe ("quiet" universe with constant entropy, radial, symmetry, microscopic masses > muonic mass)

a. th= {12.25/Jndf }(1/T**2) where ndf= number of degree of freedom = 57/4 th= Hubble's temp for evolution T=thermal temperature

b. R=Ci*(T**l/2) R = local radius

c. V=H*R H=c2 sqrt (GN)*sqrt (RHO)

152 13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

H=l/th Where Gn= Newtonian constant RHO density of gravitational matter in universe The quoted formula was adapted for a nuclear system representing nuclear matter after the first step of collisions: Gn —>ac (The Q-G constant of interaction, ac=0.128) RHO —>p (Density for nuclear matter fm~3) T—» C3 ( averaged total energy for a nuclear act, ideal gas approximation) In this way two kind of Hubble scale were defined: one of scale is reported to the conventional freeze-out point: to~l/min

< > ti ^H i b min -.2

tto—{0.01—>1, the freeze-out point} the second temporal scale is directly defined between tH=0 up to tn=t freeze-out using as unit the fm/c (lfm/c=0.33*10"23s). It was deduced introducing formulae a and b in c:

lH ~~ I n J FRZ

153 TR0700353

13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NEUTRONIC LIMITS IN VARIOUS TARGET MEDIUMS DRIVEN BY A PROTON BEAM OF 1 GEV ENERGY

Gamze GENÇ, Nesrin DEMİR, and Hüseyin YAPICI Erciyes Üniversitesi Mühendislik Fakültesi, 38039 Kayseri, Türkiye Tel: 00-90-352-437 49 01/32128 Fax: 00-90-352-437 57 84 E-mail: [email protected]

ABSTRACT The concept of accelerator-driven system (ADSs) combines a particle accelerator with a subcritical core. The basic process in an AD is nuclear transmutation, and in general, an ADS consists of three parts: (1) accelerator, (2) spallation-neutron target (SNT) and (3) sub-critical core which surrounds the SNT. This study presents the neutronic characteristics of integral data in an infinite target medium driven by an isotropic point source of 1 GeV incident proton. Lead- bismuth eutectic, mercury, tungsten, uranium, thorium, chromium, cupper and beryllium are considered as the target material because of their favorable spallation-neutron production characteristics. In order to be able to simulate the infinite target medium by eliminating the spatial dependence, a spherical target is considered, and its radius is increased gradually up to adequate radius ensuring the infinite target medium. In this way, the radius value ensuring the maximum neutron leakage out of the target would be determined. Numerical calculations were performed with the high-energy Monte Carlo code MCNPX in coupled neutron and proton mode using the LAI 50 library. The results bring out that a significant ratio (92-99%) of the leaking neutrons is below 20 MeV in the lead-bismuth eutectic, mercury, tungsten and uranium target materials. The maximum neutron leakage quantities for chromium, cupper and beryllium are an extremely low for a SNT.

Keywords: Infinite medium calculations; Spallation-neutron target; Spallation neutrons; Proton accelerator

154 TR0700354

13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

BURNUP STUDIES OF THE SUBCRITICAL FUSION-DRIVEN IN-ZINERATOR

Carl-Magnus Persson1, Waclaw Gudowski Dept. of Nuclear and Reactor Physics, Royal Institute of Technology (KTH) S-106 91 Stockholm, Sweden Francesco Venneri General Atomics San Diego, California, USA E-mail l:[email protected]

ABSTRACT A fusion-driven subcritical core, "In-Zinerator", has been proposed for nuclear waste transmutation [1]. In this concept, a powerful Z-pinch neutron source will produce pulses of 14 MeV neutrons that multiply in a surrounding subcritical core consisting of spent fuel from the LWR fuel cycle or from deep burn high temperature reactors. The proposed design has pulse frequency 0.1 Hz and a thermal power of 3 GWth. The Z-pinch fusion experiment is located at Sandia Laboratories, USA, and can today fire once a day. However, investigations have been made how to increase the frequency to several fires per minute. Each fire yields 300 MJ corresponding to 1020 neutrons per pulse. The source chamber will in the In- inerator concept be surrounded by spent fuel to reach an effective multiplicatin factor, keff, of 0.97. The core will be cooled by liquid lead. In this paper, the burnup of different fuel compositions in the In-Zinerator will be studied as function of initial keff. The Monte Carlo based continuous energy burnup code MCB [2][3]will be used.

References [1] B.B. Cipiti, Fusion Transmutation of Waste and the Role of the In-Zinerator in the Nuclear Fuel Cycle, Sandia Report SAND2006-3522, Sandia National Laboratories, USA, 2006. [2] J. Cetnar, J Wallenius and W Gudowski, MCB: A continuous energy Monte-Carlo burnup simulation code, Actinide and fission product partitioning and transmutation, Proc. of the Fifth Int. Information Exchange Meeting, Mol, Belgium, 25-27 November 1998, 523, OECD/NEA, 1998. [3] http://www.nea.fr/abs/html/nea-1643.html

155 TR0700355

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TRANSMUTATION OF HIGH LEVEL WASTES IN A FUSION-DRIVEN TRANSMUTER (FDT)

Nesrin DEMİR, Gamze GENÇ, and Hüseyin YAPICI Erciyes Üniversitesi Mühendislik Fakültesi, 38039 Kayseri, Türkiye Tel: 00-90-352-437 49 01/32107 Fax: 00-90-352-437 57 84 E-mail: [email protected]

ABSTRACT This study presents the transmutations of both the minor actinides (MAs: 237Np, 241Am, 243Am and 244Cm) and the long-lived fission products (LLFPs: 99Tc, 129I and 135Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZMA and TZpp which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fraction of the MA is raised from 10 to 20% stepped by 2%. The MAs are cladded with the graphite (10%) and cooled with the high-pressured helium gas for nuclear heat transfer. The volume fraction of helium is reduced from 80 to 70% depending on that of MA. Furthermore, the volume fraction of graphite is raised from 10 to 80% stepped by 5% to slow down the energy of neutrons entering into the TZFP while the volume fraction of LLFP is reduced from 80 to 10% depending on the graphite volume fraction. The calculations are performed for an operation period (OP) of up to 10 years by 75% plant factor (r\) under a neutron wall load (P) of 5 MW/m2 to estimate neutronic parameters and transmutation characteristics per D-T fusion neutron. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the 99Tc nuclide among them has the highest transmutation rate.

Key Words: Fusion reactors; fusion-driven transmuters; minor actinides; long-lived fission products; tritium breeding.

156 SESSION 9: PLENARY SESSION .;;:;.; zlZL-~——^~- —1~-

TR0700356

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

ADVANCED AND SUSTAINABLE FUEL CYCLES FOR INNOVATIVE REACTOR SYSTEMS

1 1 112 2 J.-P. Glatz , R. Malmbeck , D. Serrano-Purroy , P. Soucek , T. Inoue , K. Uozumi i European Commission, JRC, Institute for Transuranium Elements, Postfach 2340, 76125 Karlsruhe, Germany E-Mail: Jean-Paul. GLA TZ@ec. europa. eu 2 Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1 Iwado Kita, Komae-shi, Tokyo 201-8511, Japan ABSTRACT The key objective of nuclear energy systems of the future as defined by the Generation IV roadmap is to provide a sustainable energy generation for the future. It includes the requirement to minimize the nuclear waste produced and thereby notably reduce the long term stewardship burden in the future. It is therefore evident that the corresponding fuel cycles will play a central role in trying to achieve these goals by creating clean waste streams which contain almost exclusively the fission products. A new concept based on a grouped separation of actinides is widely discussed in this context, but it is of course a real challenge to achieve this type of separation since technologies available today have been developed to separate actinides from each other.

In France, the CEA has launched extensive research programs in the ATALANTE facility in Marcoule to develop the advanced fuel cycles for new generation reactor systems. In this so- called global actinide management (GAM) concept, the actinides are extracted in a sequence of chemical reactions (grouped actinide extraction (GANEX)) and immediately reintroduced in the fuel fabrication process is to use all actinides in the energy production process.

The new group separation processes can be derived as in this case from aqueous techniques but also from so-called pyrochemical partitioning processes. Significant progress was made in recent years for both routes in the frame of the European research projects PARTNEW, PYROREP and EUROPART, mainly devoted to the separation of minor actinides in the frame of partitioning and transmutation (P&T) studies.

The fuels used in the new generation reactors will be significantly different from the commercial fuels of today. Because of the fuel type and the very high burn-ups reached, pyrometallurgical reprocessing could be the preferred method. The limited solubility of some of the fuel materials in acidic aqueous solutions, the possibility to have an integrated irradiation and reprocessing facility with improved economics and the higher radiation stability of the molten salt media are some of the arguments in favour of pyro-reprocessing.

Adaptations of this technology exist for the treatment of both oxide and nitride fuels. The flowsheet for the treatment of nitride fuels is similar to that of metal fuel. In the case of oxides a head-end reduction step is needed. It can be performed by direct electroreduction, where the heat generating fission products are removed and the fissile materials are recovered as an alloy, which can be again directly reprocessed by electrorefining.

158 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

The present paper describes the progress made at ITU - mainly in the frame of the network projects mentioned above - in developing the grouped actinide recycling concept with a view to the sustainability goals described above for innovative reactor systems.

In the frame of these projects, reprocessing of EBRII type metallic alloy fuel with 2% of Am and 5% of lanthanides (UöoPı^o-ZrıoAmaNds.sYo.sCeo.sGdo.s) is being carried out by electrorefining at ITU. An excellent grouped separation of actinides from lanthanides (An/Ln mass ratio = 2400) had been obtained. The high of lanthanides and their possibly detrimental interaction with the cladding material implies that they must be separated. In this sense the choice of the cathode material for the actinide recovery is essential and it could be shown that aluminium is an excellent material for a pyrochemical partitioning process. The results are confirmed in conditions simulating the scaling up (multiple run) of the process, with an accumulation of Ln in the salt. One of the major goals is the minimization of actinide losses and to thereby reduce significantly the radiotoxicity of the waste produced.

The results shown here represent the first demonstration of an efficient grouped actinide recovery from realistic metallic fuels and are therefore an important step in achieving the sustainability goals of future reactor systems.

159 TR0700357 j 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

GAMMA RAY BEAM TRANSMUTATION

K. Imasaki, and D. Li, Institute for Laser Technology,2-6 Yamada-oka, Suita, Osaka, Japan Tel&Ffax: 81-6-6879-8739 E-mail: [email protected]

S. Miyamoto, S. Amano, and T. Motizuki Laboratory of Advanced Science and Technology for Industry, University of Hyogo, Japan

ABSTRACT We have proposed a new approach to nuclear transmutation by a gamma ray beam of Compton scattered laser photon. We obtained 20MeV gamma ray in this way to obtain transmutation rates with the giant resonance of I97Au and 129Iodine. The rate of the transmutation agreed with the theoretical calculation.

Experiments on energy spectrum of positron, electron and neutron from targets were performed for the energy balance and design of the system scheme. The reaction rate was about 1.5~4% for appropriate photon energies and neutron production rate was up to 4% in the measurments.

We had stored laser photon more than 5000 times in a small cavity which implied for a significant improvement of system efficiency. Using these technologies, we have designed an actual transmutation system for 129Iodine which has a 16 million year's activity. In my presentation, I will address the properties of this scheme, experiments results and transmutation system for iodine transmutation.

160 TR0700358

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

RESULTS OF THE IAEA CRP ON 'STUDIES OF ADVANCED REACTOR TECHNOLOGY OPTIONS FOR EFFECTIVE INCINERATION OF RADIOACTIVE WASTE

W. Maschek1, A. Stanculescu2, V. Gopalakrishnan3, B. Arien4, E. Malambu4, D. Da Cruz5, H. Wider6, M. Schikorrl, J. Uhlff7 ,Ch. Chabert8, Y. Wu9, A. Rineiskil, S. DulIalO , M. Szieberthll, P. Vertesl2, V. Ignatievl3, M. Vorotyntsevl4 1 Forschungszentrum Karlsruhe (FZK), P.O. Box 3640, D-76021 Karlsruhe, Germany E-mail: [email protected] international Atomic Energy Agency (IAEA), Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna, Austria 3 Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102, India 4SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol, Belgium 5Nuclear Research and Consultancy Group (NRG), P.O. box 25, 1755 ZG Petten, The Netherlands 6Joint Research Centre of the European Commission (JRC), 1755 LE Petten, The Netherlands 7Nuclear Research Institute Rez pic, 25068 Rez, Czech Republic 8 Commissariat â l'Energie Atomique (CEA) Cadarache, 13108 Saint Paul Lez Durance Cedex, France 9 Chinese Academy of Sciences (ASIPP) P.O. Box 1126, Hefei, Anhui 230031, China 10Politecnico di Torino (PT), Corso Duca degli Abruzzi 24, 10129 Torino, Italy "Budapest University of Technology and Economics (BUTE), Muegyetem rpk.9, 1111 Budapest, Hungary 12KFKI Atomic Energy Research Institute, P.O. Box 49, 1525 Budapest, Hungary 13RRC-Kurchatov Institute (RRC-KI), 123182, Kurchatov sq.l, Moscow, Russian Federation 14 Institute of Physics and Power Engineering (IPPE), 249020 Obninsk, Kaluga Region, Russian Federation

ABSTRACT The IAEA has initiated a Coordinated Research Project (CRP) on "Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste". The overall objective of the CRP, performed within the framework of IAEA's Nuclear Power Technology Development Section's Technical Working Group on Fast Reactors (TWG-FR), is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. More specifically, the final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, especially systems with high minor actinide content. Currently, 20 institutions from 15 member states and one international organization are participating in this CRP. The current author list comprises the participants of the last CRP Vienna meeting.

The CRP concentrates on the assessment of the transient behaviour of various transmutation systems. For a sound assessment of the transient and accident behaviour, neutron kinetics and dynamics methods and codes have to be qualified, especially as the margins for the safety

161 13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye relevant neutronics parameters are generally becoming small in a transmutation system. Hence, the availability of adequate and qualified methods for the analysis of the various systems is an important point of the exercise. A benchmarking effort between the codes and nuclear data used for the analyses has been performed, which will help specifying the range of validity of methods, and also formulate requirements for future theoretical and experimental research. Should transient experiments become available during the course of the CRP, experimental benchmarking work will also be pursued.

162 TR0700359 i 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

THE PROMISES AND CHALLENGES OF FUTURE REACTOR SYSTEM DEVELOPMENTS

Si-Hwan Kim, Moon Hee Chang and Hyun-Jun Kim Korea Atomic Energy Research Institute P.O. Box 105, Yusung, Daejon, KOREA 305-360 E-mail: [email protected], [email protected], [email protected]

ABSTRACT Nuclear power is an inevitable option in Korea to overcome the scarcity of national energy resources and to reduce its overseas energy dependency. During the past three decades, Korea has accomplished outstanding achievements in facilitating a nuclear power development. The share of nuclear power in electricity generation has been rapidly increasing since 1978. Nuclear power has provided Korea with a most economically and environmentally-friendly way of generating electric energy, and has contributed a lot to its national economy growth. It will continue to do so in the future. For a stable and economical supply of electricity, nationwide efforts toward achieving self-reliance in nuclear power technology have been pursued. To date, a series of nuclear technology self-reliance programs such as CANDU fuel technology, PWR fuel technology, and nuclear reactor (KSNPP) technology have been successfully completed. KSNP is a technologically advanced power plant modified by Koreas' own operating experience and domestic technology and designed by adapting several advanced technologies suitable for its national situation. The KSNP was applied to the construction of Yonggwang 5&6 and Ulchin 5&6 and is now being replicated to provide a stable, economical and reliable electric power supply. Through a comprehensive nuclear R&D programs, an enhancement of its indigenous nuclear technology capability is currently being pursued. The effort has focused on improving its indigenous nuclear power technology such as improvements in safety and economy of the KSNP (KSNP+), a 600 MWe class KSNP and advanced fuels, and the establishment of industrial codes & standards. In addition, a Korean Advanced Power Reactor (APR 1400) and a System- integrated Modular Advanced ReacTor (SMART) are currently under development.

The APR 1400 with a capacity of 1,400 MWe will be characterized by its drastically enhanced safety, reliability, and operability as well as its improved economy when compared to currently the existing plants. The APR 1400 has been developed since 1991 and it is expected that its first commercial operation will be in 2012. In the short term by 2011, the APR-1400 design will be improved from the viewpoints of safety, economics and performance. We are also developing a small integral reactor SMART, which is a promising advanced small and medium-size power category of nuclear reactors. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. SMART is purposed for dual applications such as for seawater desalination and electricity generation. Since the SMART technology is technically sound and has sufficient economics, the SMART desalination plant has good prospects of being deployed as a nuclear desalination plant.

We are also actively participating in the GEN IV collaboration (GIF: GEN IV International Forum) for a VHTR and a SFR technology development. Through close collaboration with GIF, a proliferation-resistant SFR technology will be developed based on KALIMAER for an effective uranium utilization and waste minimization. Also a high temperature reactor is currently under

163 13'h International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye development to demonstrate a nuclear based hydrogen production technology. Korea is really looking ahead by developing new generation of advanced nuclear reactor systems for a sustainable development, economical benefits, a clean environment and public confidence.

In this paper, Korean nuclear reactor technology development program is described together with lessons learned from self-reliance in nuclear reactor technology. In addition, this paper presents the status of the next generation reactor system development program and the future reactor system development program for addressing these challenges.

1970s 1980s 1990s 2000s 2010s 2020s

Introduction of Promotion of Technology Development of " Improvement of Development of Nuclear Power Localization Self-reliance Advanced Reactor : Advanced Reactor;, Future Reactor

ilplli iiSSl

Construction Establishment of OPR1000 APR 1400 APR 1400* Gen IV Systems ofKori#1 ('71-78) localization Plan ('84)' Development ('95) I Development ('01) % SMART (SFR. VHTR etc.)

' OPR1000 (Optimized Power Reactor 1,000) is new name for the former KSNP

164 PARALLEL SESSION 10A: FISSION REACTORS V TR0700360

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

STATUS AND OVERVIEW OF INPRO

Masanao Moriwaki International Atomic Energy Agency (IAEA): Wagramer Strasse 5, P.O. Box 100, A- 1400 Vienna, Austria, E-mail: [email protected]

ABSTRACT INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) was initiated in the year 2000 following a resolution of the General Conference of the IAEA in order to provide a forum for discussion of experts and policy makers on all aspects of nuclear energy planning as well as on the development and deployment of innovative nuclear energy systems (INS). It brings together technology holders, users and potential users to consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles, and it pays particular attention to the needs of developing countries. Since its initiation, INPRO has kept growing as an international project on INS, and as of August 2006 it has 26 members including major technology holder countries and technology user and potential user countries.In July 2006, INPRO had achieved a major milestone, which is finalizing Phase 1 and starting a new stage, Phase 2 with endorsement by INPRO steering committee. Phase 1 can be characterized as INPRO methodology development stage. The INPRO methodology is a methodology to assess if a given INS can fulfil demands to be a sustainable energy option in the 21st century, in a holistic way. The INPRO methodology's holistic approach includes seven areas in its sustainability assessment: Economics, Safety, Environment, Waste Management, Proliferation Resistance, Physical Protection and Infrastructure. The demands of the methodology on INS form a hierarchy consisting of Basic Principles (BPs), User Requirements (URs) and Criteria (CR). The main achievement of Phase 1 is to define BPs, URs and CR in all seven areas (summarized as TECDOC 1434) with application tests performed by INPRO Members and create a draft manual to guide how to assess INS with INPRO methodology in a practical manner. Seven chapters of the draft manual (Overviews, Economics, Safety on nuclear power plants, Safety on fuel cycle facilities, Environment, Waste Management and Infrastructure) were distributed to INPRO Members.For Phase 2, three directions are defined in our Action Plan. The first direction contains methodology oriented activities, where the methodology will be continuously improved with feedback from 12 assessment studies being performed by INPRO Members as well as publication of INPRO manual as a TECDOC. The first direction also includes creation of a vision report on opportunities and challenges of large-scale nuclear . The second direction contains institutional/infrastructure oriented activities, which will address the future infrastructure needs requiring innovation and infrastructure requirements for plants to be deployed in the future. The third direction contains collaborative project related activities, which are newly introduced in Phase 2. The collaborative projects will be performed by groups of interested INPRO Members on a variety of topics related to INS development and deployment. INPRO/IAEA will provide secretariat function for INPRO Members to plan and implement the projects. The detail framework for the collaborative projects is fully specified in a document, which was endorsed by INPRO steering committee, and now INPRO is waiting for proposals from INPRO Members.

This paper will summarize the main achievement of Phase 1 and address the status and perspective of Phase 2 with the most update information.

166 TR0700361

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ADVANCED ORIENT CYCLE, FOR STRATEGIC SEPARATION, TRANSMUTATION AND UTILIZATION OF NUCLIDES IN THE NUCLEAR FUEL CYCLE

Masaki Ozawa*13, Reiko Fujita2, Shinichi Koyama1,Tatsuya Suzuki3 and Yasuhiko Fujii3

*l Japan Atomic Energy Agency *2 Toshiba Corporation *3 Tokyo Institute of Technology, Japan E-mail: ozawa. [email protected]

ABSTRACT Electrolytic extraction (EE) method has been studied as a vital separation tool for new reprocessing process to realize transmutation and utilization of the specific fission products including LLFP* {e.g., Tc*, Ru, Rh, Pd*, Se* and Te*, etc) in the . In an employed EE process, Pd2+ cation itself would not only be easily (>99%) deposited from various nitric acid solutions, but enhance also the deposition of co-existing RuNO3+ and Re(V by acting as a catalyst (as Vâ.adatom). Such a catalytic electrolytic extraction (i.e., CEE) method was also applicable in the case of "TcCV deposition as well. Addition of Pd2+ caused either to change the dendritic metal deposition form or to improve electrochemical property of deposits. The RMFP deposit, especially quaternary-, Pd-Ru-Rh-Re, deposits on the Pt electrode obtained by the CEE method were rather spherical in shape, seemed to be electrochemically agglomerated by nano particles. The deposits were stable and showed electrochemically nobler initial hydrogen evolution potential (

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TECHNO-ECONOMIC STUDY OF HYDROGEN PRODUCTION BY HIGH TEMPERATURE ELECTROLYSIS COUPLED WITH AN EPR - WATER STEAM PRODUCTION AND COUPLING POSSIBILITIES

R. Rivera-Tinoco1. C. Mansilla2, F. Werkoff2, C. Bouallou1 1 CEP-Paris, Ecole Nationale Superieure des Mines de Paris. 60 bd Saint Michel 75006 Paris, France 2 CEA/SACLAY-DEN/DM2S/SERMA Bât 4070-91191 Gif sur Yvette CEDEX, France E-mail: [email protected]

ABSTRACT Nuclear reactors present a wide range of coupling possibilities with several industrial processes, hydrogen production being one of them. Among the Pressurised Water nuclear Reactors (PWR), the new European Pressurised Reactor (EPR) offers the water steam production at low-medium temperatures, from 230°C to 330°C for the primary and secondary exchange circuits. The use of this water steam for hydrogen production by High Temperature Electrolysis is the subject of this study, under a French context. The study of this coupling, has considered two hypotheses. First, water steam drawing off in secondary circuit has been evaluated in terms of possible impact in electricity production and reactor availability. After the drawing off at 78 bar (EPR secondary circuit pressure), pressure has to be dropped in order to protect the high temperature electrolyser from damage, so an isenthalpic drop has been considered. Liquid-vapour equilibrium happens with pressure drops, so separation of gas phase and recycling of liquid phase are proposed. Second, only water steam production with an EPR has been evaluated. The feed water enters the secondary circuit and passes from liquid phase to vapour in the steam generators, and then all steam is canalized to the high temperature electrolyser. The potentiality of water steam production in the EPR has been evaluated from 15 to 40 bar. Small reactors could be the best choice if only water steam production is considered. After steam production, it steam enters into the High Temperature Electrolysis process, like a cold stream for two parallel series of three heat exchangers reaching temperatures up to 950°C. Then the steam is heated by an electric device and finally it enters the electrolyser. The electrolysis product streams (hydrogen-steam mixture and oxygen) are used in the heat exchangers like hot streams. For both hypotheses, information about water composition has been studied in order to minimise the emission of hazard materials and electrolyser damage. Further information about electric and thermal energy production cost, electrolyser cost, heat exchangers costs, etc. has been considered and used in the techno- economic study. Concerning the electrolyser, we considered that electric needs are supplied by the electric network. An optimisation method, based on genetic algorithms has been used to estimate the lowest hydrogen production cost. Results from the optimisation method were confronted with potential steam water production, using or drawing off an EPR, to find the best coupling for hydrogen production. The drawing off of EPR secondary circuit seems to be more viable than total water production. Even pilot plant court-dated construction could be considered. Besides, the cost of 1 kilogramme of hydrogen for different water steam conditions has been estimated, being between 2.26 and 2.50 euros. This cost production seems to be near to the international goal of 2 euros.

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References

(1) Palier W-1300, Centrale de Nogent, Tranches 1-2, Region d'equipement Paris. EDF, France. December 1986

(2) L'EPR, AREVA, France. January 2006, (3) http://www.areva-np.com/scripts/info/publigen/content/templates/show.asp?P=494&L= FR&SYNC=Y&ID CAT=305, date accessed: 15/11/2006

(4) IAEA-TECDOC-1505 Data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (DAWAC) Report of a coordinated research project 2001-2005, Nuclear Fuel Cycle and Materials Section, Austria. June 2006

(5) Jon SIGURVINSSON, Christine MANSILLA et al. Heat transfer problems for the production of hydrogen from geothermal energy. Energy Conversion and Management 47 (2006)3543-3551

(6) Christine MANSILLA et al. Heat management for hydrogen production by high temperature steam electrolysis, Energy (2006), doi: 10.1016/j.energy.2006.07.033

(7) DGEMP-DIDEME. Coûts de reference de la production electrique. Secretariat d'Etat a l'lndustrie—Ministere de l'Economie, des Finances et de FIndustrie, 2003. date accessed 3/11/2006.

(8) Jon SIGURVINSSON. The production of hydrogen by High Temperature Electrolysis and Alkaline Electrolysis in a context of sustainable development. Üniversite Joseph Fourier - CEA, France. November 2005

(9) Technology Options for the Near and Long Term 2005. U.S. Climate Change Technology Program, http://www.climatetechnology.gov/library/2005/tech-options/tor2005-fullreport.pdf, date accessed: 9/11/2006

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HYBRID NUCLEAR CYCLES FOR NUCLEAR FISSION SUSTAINABILITY Mireia Piera*, Jose M. Martinez-Val, *E.T.S.I.Industriales, UNED Ciudad Universitaria, s/n -28040 Madrid-Spain E-mail: [email protected]

ABSTRACT Nuclear fission can play and must play an important role in paving the road to Energy Sustainability. Nuclear Fission does not produce CO2 emissions, and it is already exploited at commercial level with the current NPP (Nuclear Power Plants). Most of them are based on LWR reactors, which have a very good safety record. It must be noted, however, that all LWR (including the advanced or evolutionary ones) have some drawbacks, particularly their very poor efficiency in exploiting the natural resources of nuclear fuels. In this paper, an analysis is presented on how to maximize the energy actually generated from the potential contents of fission natural resources. The role of fertile-to-fissile breeding is highlighted, as well as the need of attaining a very high safety performance in the reactors and other installations of the fuel cycle. The proposal presented in this paper is to use advanced and evolutionary LWR as energy- producing reactors, and to use subcritical fast assemblies as breeders. The main result would be to increase by two orders of magnitude the percentage of energy effectively exploited from fission natural resources, while keeping a very high level of safety standards in the full fuel cycle. Breeders would not be intended for energy production, so that safety standards could rely on very low values of the thermal magnitudes, so allowing for very large safety margins for emergency cooling. Similarly, subcriticality would offer a very large margin for not to reach prompt- criticality in any event. The main drawback of this proposal is that a sizeable fraction of the energy generated in the cycle (about 1/3, maybe a little more) would not be useful for the thermodynamic cycle to produce electricity. Besides that, a fraction of the generated electricity, between 5 and 10 %, would have to be recirculated to feed the accelerator activating the neutron source. Even so, the overall result would be very positive, because more than 50 % of the natural resources could be exploited with such a cycle, using very safe reactors. This percentage is much higher than the actual value for the once-through cycle (0.5 %) and the value for multiple Pu recycling in the MOX scheme (1 %). Moreover, thorium could also be exploited through fertile conversion into U-233 in the subcritical breeders. The separation between energy production (to be done in LWR) and nuclear breeding (to be done in subcritical hybrids) presents a scenario with very appealing safety features and a high potential for an efficient utilization of all natural resources of uranium and thorium, that account for 1024 J, i.e., 25 Gtoe, which is 35,000 times as large as the annual production of Nuclear Energy nowadays, and about 2,500 times as large as the total annual energy consumption all over the globe.

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POLITICS OF NUCLEAR POWER AND FUEL CYCLE

Rizwan-uddin Department of Nuclear, Plasma and Radiological Engineering University of Illinois at Urbana-Champaign, USA E-mail: [email protected]

ABSTRACT Background It has long been recognized by those involved in the nuclear business that because of its close relationship with "the bomb," the technology will always be on the verge of controversy. Increasing energy demand will at the same time continue to propel nuclear to the center stage. However, no matter how strong the need for a safe, reliable, green and efficient source of energy—that can possibly be filled by nuclear—due to its large capital cost, potential user countries will not move down the nuclear path seriously without a consistent and long term global nuclear policy. NPT and its recent extension have provided such a framework for the last four decades. Recent political developments have however cast a cloud of uncertainty that will likely hinder countries interested in pursuing peaceful uses of nuclear power. It is imperative that the cloud of uncertainty be lifted expediently and a unified policy be developed that can provide the framework for future development of peaceful uses of nuclear power. Ad-hoc, country specific policies—dictated by proliferation concerns—are more likely to hurt non-proliferation than help. In this paper we will identify current status of nuclear development and its potential in various countries—developed and developing. Different scenarios for a unified global policy will be given and their chances of success will be discussed. It should be recognized that despite the global nature of the treaty and large number of signatories, NPT is essentially a tool in the hands of about twenty or so countries: the haves (five); non-signatories and 10-12 other signatory countries that linger on the boundary with potential aspiration for either nuclear power or even nuclear bomb. Number of countries that have nuclear power plants is now around 30. A second, slightly larger, set of countries that are directly affected by treaties like NPT are the ones in the nuclear supplier group. The list of countries in the third category—the potential aspirants—is likely to remain evolving depending on regional and global affairs. Opposition or support for nuclear technology is also likely to be a function of regional and global politics. In response to such pressures, IAEA is organizing a workshop of 140 countries to discuss proposals to guarantee countries' supply of nuclear fuel (September 19-21-, 2006; Vienna).

Premise and Question A single nuclear power plant in a country may be good for the prestige of the country, but such units are unlikely to make a major impact on the energy scene. Hence, in order for nuclear power to play a significant role, countries that decide to "go nuclear," would most likely want to diversify a significant fraction of their electricity generating capacity (and possibly heating and, in the future, hydrogen production) to nuclear, possibly requiring at least few and possibly many nuclear power plants. In order to proceed with the nuclear option, these countries would expect a certain level of long term assurance on the fuel supply. What is the kind of options that would satisfy the needs of these countries and at the same time addressing the non-proliferation concerns?

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Options The options available to countries for their nuclear program can be categorized as follows. A. Fully indigenous program with complete development of power plants and fuel cycle. B. Fully or partly indigenous program for power plant development; while depending on international consortium for fuel supply and waste treatment. C. Rely on international consortia to build and operate all aspects of nuclear power plants (with local manpower).

Others A total of around fifty to seventy five countries are likely to be interested in nuclear power in the next fifty years. These can be divided in to the three groups (A-C) given above. It is likely that, with time, there will be some expectation to move to higher levels (C to B and B to A). Countries already in group A and those willing to start in group C do not pose an issue. It is those that want to start in group A or those willing to start in group B—if appropriate assurances and guarantees are provided that are addressed in this paper. Note that under the current NPT, signatories have an "inalienable right ... to develop research, production, and use of nuclear energy for peaceful purposes without discrimination and in conformity with Articles I and II." Moreover, paragraph 2 of Article IV further underscores that each NPT state-party "undertake[s] to facilitate, and have the right to participate in, the fullest possible exchange of equipment, materials and scientific and technological information for the peaceful uses of nuclear energy." Fuel cycle is clearly a part of the peaceful uses, and hence, it is the responsibility of those concerned about proliferation to provide adequate framework and guarantees to convince countries to join group B rather than A.

Those concerned with the proliferation issue identify two major weaknesses in the NPT:

1. Ability of some signatory countries to proceed with nuclear activities hidden from IAEA oversight.

2. Concern that a country that acquires nuclear technology as a signatory can easily withdraw from the treaty and then use acquired know how for bomb making purposes.

Giving due weight to both sides of the debate—proliferation concern as well as concern that too restrictive framework may limit legitimate use of peaceful use of nuclear power—full paper will examine options for global and regional frameworks to maximize safety and fruits of beneficial uses of nuclear power.

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ANTIPROTON-INDUCED FISSION FOR SPACE POWER AND PROPULSION APPLICATIONS

Terry Kam mash University of Michigan, Ann Arbor,MI 48109, USA E-mail: [email protected]

ABSTRACT The Gasdynamic Mirror[GDM] fusion reactor is investigated for use as a bi-modal propulsion device when driven by antiprotons.The deuterium-tritium[DT] fusion reactions in the device will be initiated by the heating provided by the fission fragments and annihilation products resulting from the "at rest " annihilation of antiprotons in U-238 target nuclei.The energetic pions and muons of the proton-antiproton[or neutron]annihilation in the U-238 nucleus can heat a DT plasma to several keV during their relatively short lifetimes.The remaining heating to about 10 keV is provided by the fission fragments.Fissioning of U-238 by "at rest "annihilation of antiprotons has been shown to be 100% efficient,and the process can thus be effectively used in heating a suitable plasma to thermonuclear temperatures.With GDM as a steady state fusion reactor,and assuming certain efficiencies for the various components of the system,we calculate the energy multiplication factor"Q"needed to sustain the steady-state operation for either the "propulsive"mode or the "power-producing" mode.With the aid of a system and mission analyses,we find that approximately 3.5 micrograms of antiprotons are required to accomplish a round mission to Mars in about 59 days.A similar amount is required to initiate and sustain the power-producing mode where gigawatts of electric power may be generated.Although roughly nanograms of antiprotons are currently produced annually, it is expected that hundreds of milligrams or even several grams will be produced annually in the next decade or so when Mars missions may be contemplated.

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PRELIMINARY NEUTRON1C DESIGN OF SPOCK REACTOR: A NUCLEAR SYSTEM FOR SPACE POWER GENERATION N. Burgio*, M. Cumo+, A. Fasano+, M. Frullini+, A. Santagata*. *ENEA FIS-ION CR Casaccia, Rome - Italy +DINCE, University of Rome "La Sapienza" Rome - Italy

Corresponding Author: Massimo Frullini, DINCE, University of Rome "La Sapienza" Corso Vittorio Emanuale II n. 244, 00192 Rome, Italy. E-mail: [email protected]

ABSTRACT Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state.

Energy [«position (MoWcm3 for 1k\V of fission power)

SO 100 Ç [degree)

Fig.l: a) Keff vs synchronized control rods rotation angle. 1>) Energy deposition per k\V at steady state.

The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out.

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A-

/î-4 T'

/,(* *vi';;ı

Fig.2: a) Temperature profiles in 1 sixth symmetric section of SPOCK. b) Plaut view cross section of the same symmetric reactor element.

The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B4C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg (fig. 2b). The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth (fig. la) and a power spatial distribution (fig. 1b) were also discussed. Through the definition of an ideal reference K ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on the feasibility of the standard nuclear data sets for the design of space nuclear devices. Finally, FLUENT 6.2.16 preliminary calculations show (fig.

2a), for a hot pin fuel temperature of 1700 K, 325 kW(h were transferred to the coolant (F= 1.8

Kg/s, AT = 150 K) with the possibility to be converted in 30 kWe| by using an advanced thermoelectric converter system. Reference [1] M. Cumo, M. Frullini, A. Gandini, A. Naviglio, L. Sorabella "MAUS - 1.5 Nuclear Reactor for Space Electric Power", ICENES 2005- Bruxelles August 2005.

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PREMISES FOR USE OF FUSION SYSTEMS FOR ACTINIDE WASTE INCINERATION

Stefan TACZANOWSKI Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Cracow 30 059, Poland E-mail: [email protected] .agh .edu.pl

ABSTRACT The motivation for the present study is induction of a change in the attitude of fusion community and first of all of the respective decision makers with regard to the fission power. The aim is to convince them that admittance of any kinship of fusion to fission energy is not the greatest threat for its deployment. The true problems of fusion power lie in the physical and technological difficulties that are hindering the achievement of reliable operation and economical competitiveness of fusion reactors. It seems that the strong objections against any symbiosis of fusion with fission, which one could observe for over two decades, are based upon the ignorance of the public unaware of the common nuclear roots of both processes. They manifest themselves, among others, in the non-negligible activity to be induced in fusion devices, as a result of the exposition of construction materials to very strong fluxes of fusion (14 MeV) neutrons. The latter ones in addition, are the source of a very serious material damage in these materials. Meanwhile, most of the real difficulties fusion power is still facing can be effectively relaxed while shifting the heavy burden of sufficient production of energy to energy rich fission process.

Seeing all this, first are reminded some important problems of existing fission power that stem from the unavoidable production of Minor Actinides, distinct by undesirable physical properties (intense radioactivity, heat release, positive reactivity coefficients). Thus, in search for solutions Fusion-Driven Incineration (FDI) subcritical systems (well remote from superprompt criticality) are proposed. Next, the problems of nuclear fusion are addressed and the use of fission energy contained in actinides of spent nuclear fuel is suggested. The main advantage of that option of fusion power, /thanks to energy release from fissions/, is the prospect of a radical reduction of necessary plasma energy gain Q to levels achievable in much smaller i.e. much more economic devices. Yet, perhaps even more important advantages of the FDI system are: well homogeneous heating distribution and - first of all - reduced load of the First Wall (FW) with 14 MeV neutrons i.e. the main source of radiation damage. Simultaneously, the alpha yield from plasma to materials directly exposed to (e.g. the FW) is reduced, whereas the neutron yield attenuation reduces the gas production, DPA and the induced activity. Though instead of D-T neutrons the fission ones appear, but are much softer (below gas production thresholds) and in a much lesser number (ca.1/3). The performed calculations show that the plasma Q can be lessened to about 1 and the 14 MeV neutron yield even by a factor of ca. 30. Finally, it is emphasised that though the radiotoxicity gathered in the FDI system alone is larger than that in a fusion system free of fission waste, yet the whole radiotoxicity of the symbiotic nuclear energy system, i.e. consisted of a Fusion Driven incinerator of transuranics (Pu, Np and Am) received from associated Light Water Reactors is to be lower.

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In conclusion, the picture of hybrid option of fusion presented herewith as a means to solve the problems of both fission and fusion based nuclear energy should facilitate the development and then launching of the fusion power.

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ADVANCED ENERGY SYSTEM WITH NUCLEAR REACTORS AS AN ENERGY SOURCE

Yasuyoshi Kato, Takao Ishizuka and K. Nikitin Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology Nl-2 2-12-1 O-okayama, MeguroO-ku, Tokyo 152-8550, JAPAN Tel: +81-3-5734-3065, Fax: +81-3-5734-2959, Email: [email protected]

ABSTRACT About two-thirds of the energy generated in a light water reactors (LWRs) core is currently dissipated to the ocean as lukewarm water through steam condensers; more than half the energy in helium (He) gas turbine high temperature gas cooled reactors (HTGRs) is dissipated through pre-coolers and intercoolers. The new waste heat recovery system 1) efficiently recovers the waste heat from reactors using boiling heat transfer of 20°C liquid carbon dioxide (CO2) instead of conventional sea water as a cooling medium. The CO2 gasified in the cooling process is used directly as a working fluid of mechanical heat pumps for hot water supply. In LWRs, the net energy utilization fraction to total heat generation in the core exceeds 85% through the waste heat recovery. This cogeneration system is about 2.5 times more effective than current systems in reducing global warming gas emissions and long half- life radioactive material accumulation. It also increases uranium resource utilization relative to current LWRs. In the HTGR cogeneration system, the waste heat is also useful for cold water supply by introducing an adsorption refrigeration system since the gas temperature is still as high as about 190°C. When the heat recovery system is incorporated into the HTGR, the electricity to heat-supply ratio of the HTGR cogeneration system accommodates the demand ratio in cities well; it would be suited to dispersed energy sources. The heat supply cost is expected to be lower than those of conventional fossil-fired boilers beyond operation of about four years. The waste heat recovered is able to be utilized not only for local heat supply but also for methane and methanol production from waste products of cities and farms through high-temperature fermentation, e.g., garbage, waste wood and used paper that are produced in cities, along with excreta produced through fanning. The methane and methanol can be used to generate hydrogen for fuel cells. The new waste heat recovery system is also applicable to a fast reactor (FR) with a supercritical CO2 gas turbine2) that achieves higher cycle efficiency than conventional sodium cooled FRs with steam turbines. The FR will eliminate problems of conventional FRs related to safety, plant maintenance, and construction costs. The FR consumes efficiently trans-uranium elements (TRU) produced in light water reactors as fuel and reduce long-lived radioactive wastes or environmental loads of long term geological disposal).

An Advanced Energy System (AES) with nuclear reactors as an energy source has been proposed which supply electricity and heat to cities. The AES has three objectives:

1. Save energy resources and reduce green house gas emissions, attaining total energy utilization efficiency higher than 85% through waste heat recovery and utilization. 2. Foster a recycling society that produces methane and methanol for fuel cells from waste products of cities and farms.

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3. Consume TRU produced in LWRs as fuel for FRs, and reduce long-lived radioactive wastes or environmental loads of long term geological disposal.

References 1. Y. Kato, T. Nitawaki and K. Fujima, "Zero Waste Heat Release Nuclear Cogeneration System," Proc. 2003 Intl. Congress on Advanced Nuclear Power Plants (ICAPP '03), Cordoba, Spain, May 4-7, 2003, Paper #3313. 2. Y. Kato, T. Nitawaki and Y. Muto, "Medium Temperature Carbon Dioxide Gas Turbine Reactor," Nucl. Eng. Design, 230, pp. 195-207 (2004). 3. H. N. Tran and Y. Kato, "New 237Np Burning Strategy in a Supercritical CO2 Cooled Fast Reactor Core Attaining Zero Burnup Reactivity Loss," Proc. American Nuclear Society's Topical Meeting on Reactor Physics (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14,2006.

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THE GDT-BASED FUSION NEUTRON SOURCE AS DRIVER OF A MINOR ACTINIDES BURNER

K. Noack, A. Rogov, A. A. Ivanov*, E. P. Kruglyakov*, Yu. A. Tsidulko* Forschungszentrum Dresden-Rossendorf, Postfach 51 01 19, D-01314 Dresden, Germany Budker Institute of Nuclear Physics, Lavrentyev Prospect 11, 630090 Novosibirsk, Russia E-mail: [email protected]

ABSTRACT To become a long-term sustainable option for the world's energy supply fission reactor tech- nology must minimize its high-level waste, which finally has to be disposed. To solve the problem, worldwide great R&D effort is made to develop new closed fuel cycle options. Long- lived fission products and, in particular, minor actinides are the components of the spent nuclear fuel which cause the most concern. Regarding the incineration of minor actinides, systems producing and confining the high-energetic (fast) neutrons have the highest effi-ciency. These systems can be built as fast reactors and as sub-critical nuclear fuel systems, the so-called driven systems, which are fed with neutrons from an outer neutron source. At pre-sent, the accelerator driven spallation neutron source is favored for this purpose thanks to the high neutron emission intensity achievable. Compared to fast reactors the combined accelera-tor driven system (ADS) has several advantages. The most important are the higher possible burning efficiency and the enhanced inherent safety characteristics. Therefore this develop-ment line is intensively pursued by several research projects, e.g. by the project EUROTRANS of the European Union [1].

The Budker Institute of Nuclear Physics Novosibirsk has made the proposal of a powerful 14 MeV neutron source on the base of the gas dynamic trap (GDT) plasma device [2,3]. This neutron source is primarily thought for an irradiation test facility of materials that must be de- veloped for the fusion DEMO reactor. A research project of the Budker Institute aims at com- pleting the database of the GDT in the range of high plasma parameters, which are relevant for the neutron source, and at demonstrating its feasibility and suitability by a hydrogen-prototype [4]. The situation outlined before raises the questions whether the GDT based neutron source could also be a candidate for driving a sub-critical system devoted to nuclear waste transmuta- tion and how this option compares to the spallation based ADS? The answers on these ques-tions are the objective of the present paper.

By means of the Monte Carlo code MCNP-4C2 and the nuclear data library JENDL-3.3 neu-tron transport calculations were carried out for a minor actinides burner the scheme of which originally has been defined as a numerical benchmark exercise for spallation based ADS by the OECD NEA [5] and afterwards was slightly modified in Ref. [6]. Basic neutron character-istics of the system were calculated for the cases when operated with both the spallation source and the GDT neutron source. The results make clear what are the differences between both cases regarding the neutronics and what they have in common. From the obtained calcu-lation results and from known parameters of both neutron sources the following main conclu-sions can be drawn: • The projected parameters of the GDT-based neutron source make it a candidate for a driver of an actinides burner. • To become competitive with the ADS, the further R&D effort for the GDT neutron source project must, above all, be directed towards an increase of its energetic efficiency.

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References [1] EUROTRANS - Integrated Project of the Sixth Framework Programme Euratom, http://www.fzk.de/eurotrans/ [2] A.A. Ivanov and D.D. Ryutov, Nucl. Sci. Eng., 106, 235 (1990) [3] P.A. Bagryansky, et al., Fusion Eng. Des., 70, 11 (235) [4] E.P. Kruglyakov, Proc. of Int. Conf. On Open Plasma Conf. Sys., Novosibirsk, A. Ka-bantsev (ed.), IAEA, 349 (1993) [5] Nuclear Energy Agency, NEA/NSC/DOC (2001)13 [6] G. Aliberti, et al., Nucl. Sci. Eng., 146, 13 (2004).

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MONTE CARLO CALCULATION FOR DIFFERENT ENRICHMENT LITHIUM MODERATOR IN A HYBRID REACTOR

Hacı Mehmet ŞAHİN3, Şenay YALÇIN", Taner ALTINOK0 and Adem ACIRb

aGazi Üniversitesi, Teknik Eğitim Fakültesi, Teknikokullar, Ankara 06503, Türkiye Office phone + Fax: 00-90-312-212 43 04 E-Mail: [email protected] [email protected]. tr bBahçeşehir Üniversitesi,Mühendislik Fakültesi,BeŞİktaş, İstanbul, Türkiye cKara Harb Okulu, Savunma Bilimleri Enstitüsü, ,Ankara 06654, Türkiye

ABSTRACT In general, the fusion-fission (hybrid) is a combination of the fusion and fission processes. In this concept, the fusion plasma is surrounded with a blanket made of the fertile materials to convert them into fissile materials by transmutation through the capture of the high yield fusion neutrons. Figure 1 shows the basic structure of the hybrid blanket adapted from previous studies to this work. A line neutron source in a cylindrical cavity simulates the fusion plasma chamber. The latter is surrounded by a first wall (FW). The stainless steel type of SS304 is used as the FW and fuel cladding material. The fissile zone is composed of natural uranium dioxide (UO2) in hexagonal geometry as 10 rows. Then, the radial reflector is made of Lİ2O for production of tritium (T) and graphite in a sandwich structure. This measure reduces the neutron leakage drastically and leads to a better neutron economy

The purpose of this work is to investigate the effect of natural lithium and different enrichment lithium between 10% and 90% utilization as a moderator, which is the nuclear heat transfer out of the fuel zone and also contributing of the tritium breeding ratio (TBR), for the neutronic parameter such as tritium breeding capability in the blanket and radiation damage; displacement per atom (DPA) and He-production (n,a) in the FW as a lifetime of 1 full power year (FPY). Neutronic calculations has been performed by recent Monte Carlo Neutron-Particle Transport code MCNP5 version 1.40 for a 14.1 MeV (D,T) fusion driver under a neutron wall loud of 2.25 MW/m2(10I4n/s).

For a self sustaining fusion reactor, a TBR > 1.05 will be required. The TBR values have been calculated in the range between 1.153 and 1.295 for natural Li-6 and 90% Li-6 enrichment, respectively. While the TBR value increases with Li-6 enrichment, the best performance TBR value of 1.295 is achived with 90% Li-6. A material demage criteria of the fusion reactor structural should be concidered both the DPA and helium production limit. In this study, a conservative radiation damage limit of 100 DPA and 500 appm are selected for DPA and helium gas production, respectively. At the FW, after 1 year operation period as DPA/FPY, the highest value 28.99 was found for natural Li-6 while the lowest value 26.31 was for 90% Li-6. In addition, the highest He-production value was found 261.21 appm/FPW for natural Li-6, whereas low values were found 258.53 appm/FPW for 90% Li-6. In this respect, first wall structure will be replaced every 1.92 and 1.94 years for natural Li-6 and 90% Li-6, respectively.

184 13lh International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

Top view

Right view

i O

12 16

©: First wall ©: Fuel ®: Li O ©-.Carbon Dimensions are given in cm 2 (not in scale)

Coolant Clad Fuel Hexagonal lattice with pitch length=1.25 cm

Figure 1. Cross sectional view of the investigated blank

185 TR0700371

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

Linear Pulse Motor Type Control Element Drive Mechanism for the Integral Reactor

Je-Yong Yu, Suhn Choi, Ji-Ho Kim, Hyung Huh and Keun-Bae Park

Power Reactor Development Center, Korea Atomic Energy Research Institute P.O. BOX 105, Yuseong, Daejeon, 305-600, Korea Phone: 82-42-868-2835, Fax: 82-42-868-8990 E-mail: [email protected],kr

ABSTRACT The integral reactor SMART currently under development at Korea Atomic Energy Research Institute is designed with soluble boron tree operation and use of nuclear heating for reactor start- up. These design features require the Control Element Drive Mechanism (CEDM) for SMART to have fine-step movement capability as well as high reliability for the fine reactivity control.

In this paper, design characteristics of a new concept CEDM driven by the Linear Pulse Motor (LPM) which meets the design requirements of the integral reactor SMART are introduced. The primary dimensions of the linear pulse motor are determined by the electro-magnetic analysis and the results are also presented. In parallel with the electro-magnetic analysis, the conceptual design of the CEDM is visualized and checked for interferences among parts by assembling three dimensional (3D) models on the computer. Prototype of LPM with double air-gaps for the CEDM sub-assemblies to lift 100kg is designed, analysed, manufactured and tested to confirm the validity of the CEDM design concept. A converter and a test facility are manufactured to verify the dynamic performance of the LPM. The mover of the LPM is welded with ferromagnetic material and non-ferromagnetic material to get the magnetic flux path between inner stator and outer stator.

The forces of LPM predicted by analytic model have shown good agreement with experimental results from the prototype LPM. It is found that the LPM type CEDM has high force density and simple drive mechanism to reduce volume and satisfy the reactor operating circumstances with high pressure and temperature.

Key words: Integral Reactor, CEDM, Linear Pulse Motor

186 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

NEUTRONIC AND THERMAL HYDRAULIC ASSESSMENT OF FAST REACTOR COOLING BY WATER OF SUPER CRITICAL PARAMETERS

Yu.D.Baranaev*, A.P.Glebov *, V.F.Ukraintsev**, V.V.Kolesov**

*- Institute of Physics and Power Engineering (FEI), Obninsk, Russia ** Obninsk Technical University for Nuclear Power Engineering (IATE), Obninsk. Russia E-mail:

ABSTRACT Necessity of essential improvement of competitiveness for reactors on light water determines development of new generation power reactors on water of super critical parameters.

The main objective of these projects is reaching of high efficiency coefficients while decreasing of investment to NPP and simplification of thermal scheme and high safety level. International programme of IV generation in which super critical reactors present is already started.

In the frame of this concept specific Super Critical Fast Reactor with tight lattice of pitch is developing by collaboration of the FEI and IATE.

In present article neutronic and thermal hydraulic assessment of fast reactor with plutonium MOX fuel and a core with a double-path of super critical water cooling is presented (SCFR-2X). The scheme of double path of coolant via the core in which the core is divided by radius on central and periphery parts with approximately equal number of fuel assemblies is suggested. Periferia part is cooling while downcoming coolant movement. At the down part of core into the mix chamber flows from the periphery assemblies joining and come to the inlet of the central part which is cooling by upcoming flow.

Eight zone of different content of MOX fuel are used (4 in downcoming and 4 in upcoming) sub zones. Calculation of fuel burn-up and approximate scheme of refueling is evaluated.

Calculation results are presented and discussed.

187 TR0700373

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

SIMULATING A PARTIAL LOCA IN A NARROW CHANNEL USING THE DSNP SIMULATION SYSTEM

D. Saphier Soreq Nuclear Research Centre, Yavne 81800, Israel; E-mail: [email protected]; Tel +97289451743

ABSTRACT A partial LOCA accidentt in a pool type research reactor was investigated. A new MTR type fuel channel model for the DSNP simulation system was developed; permitting detailed axial and radial temperature distribution. New and older heat transfer correlations were incorporated in the model. Simulation for water levels of 14 and 35 cm in a 62cm channel were performed. The resulting maximum temperatures remain significantly below the aluminium melting point, and no damage to the core will take place under these conditions.

188 SESSION 11: PLENARY SESSION TR0700374 |

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

EURATOM STRATEGY TOWARDS FUSION ENERGY

Carlos Varandas Full Professor of Instituto Superior Tecnico Chairman of the Steering Committee of the European Fusion Development Agreement, Portugal E-mail: [email protected]

ABSTRACT Research and development (R&D) activities in controlled have been carried out since the 60's of the last century aiming at providing a new clean, powerful, practically inexhaustive, safe, environmentally friend and economically attractive energy source for the sustainable development of our society.The EURATOM Fusion Programme (EFP) has the leadership of the magnetic confinement R&D activities due to the excellent results obtained on JET and other specialized devices, such as ASDEX-Upgrade, TORE SUPRA, FTU, TCV, TEXTOR, CASTOR, ISTTOK, MAST, TJ-II, W7-X, RFX and EXTRAP. JET is the largest tokamak in operation and the single device that can use deuterium and tritium mixes. It has produced 16 MW of fusion power, during 3 seconds, with an energy amplification of 0.6. The next steps of the EFP strategy towards fusion energy are ITER complemented by a vigorous Accompanying Programme, DEMO and a prototype of a fusion power plant. ITER, the first experimental fusion reactor, is a large-scale project (35-year duration, 10000 MEuros budget), developed in the frame of a very broad international collaboration, involving EURATOM, Japan, Russia Federation, United States of America, Korea, China and India. ITER has two main objectives: (i) to prove the scientific and technical viability of fusion energy by producing 500 MW, during 300 seconds and a energy amplification between 10 and 20; and (ii) to test the simultaneous and integrated operation of the technologies needed for a fusion reactor. The Accompanying Programme aims to prepare the ITER scientific exploitation and the DEMO design, including the development of the International Fusion Materials Irradiation Facility (IFMIF). A substantial part of this programme will be carried out in the frame of the Broader Approach, an agreement signed by EURATOM and Japan. The main goal of DEMO is to produce electricity, during a long time, from nuclear fusion reactions. The prototype of a fusion power plant should demonstrate the economically viability of fusion as a commercial energy technology .The time needed to achieve the final goal of fusion R&D is strongly dependent on how the above mentioned steps will be implemented. The so-called Fast Track recommends implementation as much as possible in parallel instead of in series. The current foresights are 40 to 50 years after the beginning of ITER construction.

190 TR0700375

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

FAST IGNITION STUDIES AT OSAKA UNIVERSITY

Kazuo A. Tanakai)2) and FI project team2) ^Graduate School of Engineering & Institute of Laser Engineering, Osaka University ^Institute of Laser Engineering, Osaka University Suite, Osaka 565-0871 Japan E-mail: katanaka@ile. osaka-u. ac.jp

ABSTRACT After the invention of the chirped pulse amplification technique [1], the extreme conditions of matters have become available in laboratory spaces and can be studied with the use of ultra- intense laser pulse (UILP) with a high energy. One such example is the fast ignition [2] where UILP is used to heat a highly compressed fusion fuel core within 1-10 pico-seconds before the core disassembles. It is predicted possible with use of 50-100 kJ lasers for both imploding the fuel and heating [2] to attain a large fusion gain. Fast ignition was shown to be a promising new scheme for laser fusion [3] with a PW (= 10'5 W) UILP and GEKKO XII laser systems at Osaka. Many new physics have been found with use of UILP in a relativistic parameter regime during the process of the fast ignition studies. UILP can penetrate into over-dense plasma for a couple hundred microns distance with a self-focusing and relativistic transparency effects. Hot electrons of 1-100 MeV can be easily created and are under studies for its spectral and emission angle controls. Strong magnetic fields of 10's of MGauss are created to guide these hot electrons along the target surface [4]. Based on these results, a new and largest UILP laser machine of 10 kJ energy at PW UILP peak power is under construction to test if we can achieve the sub-ignition fusion condition at Osaka University. The machine requires challenging optical technologies such as large size (0.9 m) gratings, tiling these gratings for UILP compression; segmenting four large UILP beams to obtain diffraction limited focal spot. We would like to over-view all of these activities.

References [1]D. STRICKLAND and G. MOUROU, Opt. Commun., 56, 219 (1985) [2] S. ATZENI et al, Phys Plasmas, 6, 3316 (1999) [3] R. KODAMA, K.A. TANAKA et al., Nature, 418, 933 (2002) [4] A.L. LEI, K.A. TANAKA et al., Phys. Rev. Lett., 96, 255006(2006); H. HABARA, K.A. TANAKA et al., Phys. Rev. Lett., 97, 095004 (2006).

191 TR0700376

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

EXERGY ANALYSIS OF A SYSTEM USING A CHEMICAL HEAT PUMP TO LINK A SUPERCRITICAL WATER-COOLED NUCLEAR REACTOR AND A THERMOCHEMICAL WATER SPLITTING CYCLE

Mikhail Granovskii, Ibrahim Dincer, Marc A. Rosen and Igor Pioro1 Faculty of Engineering and Applied Science, University of Ontario Institute of Technology School of Energy Systems and Nuclear Science, University of Ontario Institute of Technology 2000 Simcoe Street North, Oshawa, Ontario, Canada, L1H 7K4 E-mails: [email protected]; [email protected]; [email protected]

ABSTRACT The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625°C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%.

Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis.

However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the "nuclear" heat and, thus, the intensity of the heat transfer to the water splitting cycle.

A preliminary exergy analysis of the proposed heat pump is conducted and a coefficient of performance (COP), taking into account a decrease in electricity generation in the nuclear power generation cycle, is evaluated. The calculated per unit mass flow heat supply rate to the thermochemical cycle increases by a factor of 3 to 5 depending on the steam temperature. A combined system comprising a SCW nuclear plant, a chemical heat pump and a lower temperature UT-3 thermochemical water splitting cycle is presented. Despite a decrease in electricity generation, the calculated exergy efficiency of hydrogen production in the considered system appears to be competitive with that for low temperature water electrolysis.

192 TR0700377

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

LASER ENHANCED RADIOACTIVE DECAY AND SELECTIVE TRANSMUTATION OF NUCLEAR WASTE

Rainer Salomaa, Pertti Aarnio, Jarmo Ala-Heikkilâ, Antti Hakola, and Marko Santala Advanced Energy Systems Helsinki University of Technology POBox 4100, FIN-02015 TKK, Finland E-mail: [email protected]

ABSTRACT We have investigated narrow-band coherent laser radiation - ranging from visible to X- and to gamma-ray wave length region - and their interactions both directly with photon-nuclear couplings and indirectly through the photon-electron and electron-nucleus interactions. In particular we discuss various means of selective excitation of nuclear resonance states by narrow- band lasers. During the relaxation process the active nucleus may return to its initial ground-state or find another final state. In the latter case the nucleus is transmuted into a state which may have beneficial properties for instance concerning radioactivity. One ideal case would be the destruction of long-lived nuclear waste isotopes into faster decaying ones. The essential presumption is that the excitation process is selective and efficient as regards background processes due to unwanted excitation channels of the primary isotope and due to other surrounding nuclides. The paper consists of 1) a short review of generating short-wave length coherent light sources, 2) a survey of potential photon-induced nuclear states and their decay channels, and 3) a determination of the selectivity of the transmutation process.

193

PARALLEL SESSION 12A: NUCLEAR FUTURE TR0700378

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

RECENT PROGRESS IN STOCHASTIC TRANSPORT THEORY

A. Ziya Akcasu 2919 Cooley Building, 2355 Bonisteel Boulevard, Ann Arbor, Michigan 48109, USA E-mail: [email protected]

ABSTRACT A transport equation for the mean flux in spatially random media is derived, and is referred to as the Modified-Levermore-Pomraning equation (M-L-P). It differs from the conventional L-P equations in that | \x | in the latter is replaced by \x in M-L-P. It is shown that when scattering is present the L-P equations are always incorrect in the sense there is not any special situation in which they can lead to an exact result. In particular they always predict the relaxation lengths of the spatial modes incorrectly. On the other hand, the M-L-P equations are exact when the flux at the origin is deterministic, as in some special cases such as half infinite medium, and infinite medium with a localized source at the origin, when the density of the medium is spatially random. However, the M-L-P equations become approximate when the medium is a finite slab because of the right boundary condition. But the relaxation rates of the spatial modes are always calculated exactly even in finite slab. The nature of approximation inherent in the M-L-P is elucidated by comparison with the exact "stochastic transition matrix formalism" developed earlier in two-stream transport.

196 Ill TR0700379

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

INNOVATIONS ON NUCLEAR ENERGY - WHAT CAN A SMALL COUNTRY CONTRIBUTE?

Borut Smodis, Milan Cercek Jozef Stefan Institute, Jamova 39, SI-1000 Ljubljana, Slovenia E-mail: [email protected]

ABSTRACT The Slovenian energy policy gives priority to the use of renewable energy resources. The energy policy defined in the Resolution on the National Energy Program adopted in 2004 foresees increasing of renewable energy (RES) sources in the primary energy consumption up to 12 % in 2010. The share of electricity from RES in total electricity production in the year 2004 was 29,1%. The share of RES in the primary energy balance in the same year was 10,7 %, with about half of this coming from hydropower. The electricity power produced by co-generation is about 8 % and expected to double by 2010. However, the Krsko nuclear power plant produces about one third of electricity needed within the country. Consequently, Slovenia has long tradition in research pertaining to development and utilisation of nuclear technology. Furthermore, Slovenian scientists have long been collaborating in numerous fusion-related projects. The major equipment available include an ion-beam accelerator with material diagnostics installations, the TRIGA nuclear research reactor, high-temperature furnaces, an advanced, dedicated fully-integrated high-resolution microscope facility for investigations of nanostructured materials, computer systems for simulations, structural mechanical analysis and CAD, and much more. The researchers at the Jozef Stefan Institute study the processes that occur on plasma facing materials and in the edge plasma of tokamak reactors and involve neutral hydrogen/deuterium molecules. These molecules are typically vibrationally excited that influences respective reaction cross- sections. Therefore, a special experimental technique for vibrational spectroscopy of molecules was developed and an ion beam analytical technique ERDA is used for characterizing hydrogen content on and beneath the material surface. Ion beam analytical methods are also being developed for the studies of plasma wall interaction processes such as erosion, deposition, fuel retention and material migration in fusion reactors. Precise pressure-gauge measurements and quadrupole mass spectrometry of the up-taken and released gases are utilised in a study aimed at quantitative characterising the kinetics of deuterium interaction with reactor wall material. A group of materials scientists is focused on ceramic processing for the production of low- activation SiC-based composites as a potential material that will meet requirements for the structural application in the first-wall blanket. Neutron-photon transport calculations, the evaluation of attenuation factors and neutron in-scattering effects within the JET-diagnostics upgrade project are performed. Several types of blanket modules for ITER are being developed within the European fusion programme. Within these investigations, the sensitivity/uncertainty pre-analysis of a mock up of the Test Blanket module based on HCLL concept is performed in order to assess the uncertainty on tritium production ratio due to uncertainty in the basic nuclear data. Deterministic transport codes, special sensitivity/uncertainty code package and libraries are used for this work. Sensitivity profiles and nuclear data uncertainties, which will be determined for the neutron responses, will be used to guide and optimise the design of the benchmark experiment. Beside the research aimed at supporting the development of the JET and ITER, the Jozef Stefan Institute also takes part in the design of more "distant" fusion power plant reactors. The main objective of collaboration within the DEMO Working Group is contribution to the conventional nuclear power plant technology, particularly in the topics related to nuclear safety

197 131h International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye and nuclear waste treatment. Within this context, categorization of activated material, prepared as the basis for various power plant conceptual design alternatives, is being investigated.

198 TR0700380

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

NUCLEAR POWER AS A NECESSARY OPTION, ALBEIT AN INSUFFICIENT ONE

Prof. Dr. Vural Altın Bilim ve Teknik, Tübitak, Türkiye E-mail: [email protected]

ABSTRACT In this presentation a comparative assessment of known energy resources are made with respect to their energy densities. Fossil fuels have formed the foundation of a worldwide economic development realized throughout the 20th century. Their comparatively high energy densities have made faster energy flows and thereby higher power levels and speedy development possible. However, renewable sources that are already feasible have much lower levels of energy densities. Their large scale utilization in lieu of fossil fuels would necessitate either reduction of economic growth rates to 'sustainable' levels or speedy development of feasible large scale storage technologies. Nuclear energy appears to impose itself as a necessity to alleviate this transition period, albeit within the constraint of known uranium reserves an insufficient one.

199 TR0700381

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TOWARDS A NEW WORLD: THE CONTRIBUTIONS OF NUCLEAR ENERGY TO A SUSTAINABLE FUTURE

Romney B. Duffey*, A. I. Miller, P. J. Fehrenbach, S. Kuran, D. Tregunno and S. Suppiah * Atomic Energy of Canada Limited, Chalk River, ON, Canada KOJ 1J0 *E-mail: [email protected]

ABSTRACT Over the last few years, the world has seen growing concern about the sustainability of the Planet when supplying increasing energy use. The major issues are: increased energy prices in the world markets; growing energy demand in emerging economies; security and stability of oil and gas supply; potentially adverse climate change due to carbon-based emissions; and the need to deploy economic, sustainable and reliable alternates. Large undefined "wedges" of alternate energy technologies are needed.

In light of these major difficulties, there is renewed interest and need for a greater role for nuclear energy as a safe, sustainable and economic energy contributor. The shift has been, from being viewed by some as politically discounted, to being accepted as absolutely globally essential. We have carefully considered, and systematically, extensively and technically analyzed the contributions that nuclear energy can and should make to a globally sustainable energy future. These include restraining emissions, providing safe and secure power, operating synergistically with other sources, and being both socially and fiscally attractive. Therefore, we quantify in this paper the major contributions:

a) The reduction in climate change potential and the global impact of future nuclear energy deployment through emissions reduction, using established analysis tools which varying the plausible future penetration and scale of nuclear energy. b) The minimization of economic costs and the maximization of global benefits, including investment requirements, carbon price implications, competitive market penetration, and effect of variable daily pricing. c) The introduction of fuel switching, including base-load nuclear energy synergistically enabling both hydrogen production and the introduction of significant wind power. d) The management and reduction of waste streams, utilizing intelligent designs and fuel cycles that optimize fuel resource use and minimize emissions, waste disposal requirements and concerns. e) The available technologies to implement these strategies, their future potential and further development and deployment.

We present the steps being taken to realize this emerging nuclear potential, showing the beneficial impacts both in the short and the long term. The deployment of a large amount of nuclear energy is shown to be achievable, sensible and supportive of global sustainability.

Keywords: Energy Demand and Supply; Global Climate Change; Advanced Systems; Nuclear Energy; Sustainable Energy; Hydrogen; Fuel Resources; Wind Power; Economics.

200 PARALLEL SESSION 12B: RADIATION EFFECTS TR0700382

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

RADIATION SHIELDING PERFORMANCE OF SOME CONCRETE

I. Akkurt**, H. Akyildinm1, B. Mavi1, S. Kılmçarslan2, C. Başyigit2

Süleyman Demirel Üniversitesi, Fen-Edebiyat Fakültesi Fizik Böl. İsparta, Türkiye. 2Süleyman Demirel Üniversitesi, Teknik Eğitim Fakültesi Yapı Eğt. Böl. İsparta, Türkiye *E-mail: [email protected]

ABSTRACT The energy consumption is increasing with the increased population of the world and thus new energy sources were discovered such as nuclear energy. Besides using nuclear energy, nuclear techniques are being used in a variety of fields such as medical hospital, industry, agriculture or military issue, the radiation protection becomes one of the important research fields. In radiation protection, the main rules are time, distance and shielding. The most effective radiation shields are materials which have a high density and high atomic number such as lead, tungsten which are expensive. Alternatively the concrete which produced using different aggregate can be used. The effectiveness of radiation shielding is frequently described in terms of the half value layer (HVL) or the tenth value layer (TVL). These are the thicknesses of an absorber that will reduce the radiation to half, and one tenth of its intensity respectively. In this study the radiation protection properties of different types of concrete will be discussed.

202 TR0700383

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

TRANSMUTATION OF TECHNETIUM INTO STABLE RUTHENIUM IN HIGH FLUX CONCEPTUAL RESEARCH REACTOR

N. Amrani and A. Boucenna Physics Department, Faculty of Sciences, UFAS University, Setif 19000, Algeria E-mail: [email protected]

ABSTRACT The effectiveness of transmutation for the long lived fission product technetium-99 in high flux research reactor, considering its large capture cross section in thermal and epithermal region is evaluated. The calculation of Ruthenium concentration evolution under irradiation was performed using ChainSolver 2.20 code. The approximation used for the transmutation calculation is the assumption that the influence of change in irradiated materials structures on the reactor operator mode characteristics is insignificant. The results on Technetium transmutation in high flux research reactor suggested an effective use of this kind of research reactors. The evaluation brings a new concept of multi-recycle Technetium transmutation using HFRTRAN (High Flux Research Reactor for Transmutation).

203 TR0700384

13 International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

ENCAPSULATING OF HIGH-LEVEL RADIOACTIVE WASTE WITH USE OF PYROCARBON AND SILICON CARBIDE COATINGS

A. Chernikov SIA - Lutch Podolsk, Moscow Region, 142100, Russia E-mai 1: chernikov@sialuch. ru

ABSTRACT It is known that high-level radioactive waste (HLW) constitute a real danger to biosphere, especially that their part, which contains transuranium and long-lived radionuclides resulting during reprocessing of nuclear fuel industrial and power reactors. Such waste contains approximately 99 % of long-lived fission products and transplutonium elements. At present, the concept of multibarrier protection of biosphere from radioactive waste is generally acknowledged. The main barriers are the physicochemical form of waste and enclosing strata of geological formation at places of waste-disposal. Applied methods of solidification of HLW with preparation of phosphatic and borosilicate glasses do not guarantee in full measure safety of places of waste-disposal of solidified waste in geological formations during thousand years. One promising way to improve HLW handling safety is placing of radionuclides in mineral-like matrixes similar to natural materials. The other possible way to increase safety of HLW disposal places is suggested for research by experts of Russian research institutes, for example, in the proposal for the Project of ISTC and considered in the present report, is to introduce an additional barrier on a radionuclides migration path by coating of HLW particles. Unique protective properties of pyrocarbon and silicon carbide such as low coefficients of diffusion of gaseous and solid fission products and high chemical and radiation stability [1] attract attention to these materials for coating of solidified HLW. The objective of the Project is the development of method of HLW encapsulating with use of pyrocarbon and silicon carbide coatings. To gain this end main direction of researches, including analysis of various encapsulation processes of fractionated HLW, and expected results are presented. Realization of the Project will allow to prove experimentally the efficiency of the proposed approach in the solution of the problem of HLW conditioning and ecological safety of waste-disposal in geological formations. References (1) R.J. Price. Properties of silicon carbide for nuclear fuel particle coatings. (2) E. Groos, G. Mielken et al. Fission product release from coated particles embedded in spherical fuel elements for high-temperature reactors. - J. Nucl. Technology, 1977, vol. 35, No. 2, pp. 320 - 326; pp. 509-515.

204 TR0700385

13 International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

DETERMINATION OF SHIELDING PARAMETERS FOR DIFFERENT TYPES OF CONCRETES BY MONTE CARLO METHODS

J)A. Aminian and 2)M. R. Nematollahi Department of Nuclear Engineering, Shiraz University, Zand Ave., Shiraz, Iran E-mail1 :[email protected] E-mail :nema@shirazu. ac. ir

ABSTRACT The chose of a suitable concrete composition for a biological reactor shield remain as a research target up to now. In the present study the attempts has been made to estimate the influence of the concrete aggregates on the shielding parameters for three type of ordinary, serpentine and steel- magnetite concrete by Monte Carlo N-Particle (MCNP ) transport code. MCNP calculations have been performed in order to obtain the leakage of neutrons, photons and electrons from dry shield. Also the mass attenuation coefficients and the liner attenuation coefficient are estimated for neutron and photon in those energies in range of actual energy which exist out of pressure vessel of power reactor in the cavity for the investigated concretes. The concrete densities ranged from 2.3 to 5.11 g/cm3. These calculations were done in the condition of a typical commercial Pressurized Water Reactor (PWR). The results show that Steel-magnetite concrete, with high density (5.11 g/cm3) and constituents of relatively high atomic number, is an effective shield for both photons and neutrons.

Keywords: Dry shield, MCNP, PWR, Serpentine Concrete & Steel-magnetite Concrete

205

PARALLEL SESSION 12C MISCELLANEOUS TR0700386

13 International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PURPOSEFUL SYNTHESIS OF CHEMICAL ELEMENTS AND ECOLOGICALLY PURE MOBILE SOURCES OF ENERGY

Vladimir A. Krivitsky* and Fangil A. Gareev** * Dubna University, Dubna, Russia ** Joint Institute for Nuclear Research, Dubna, Russia E-mail: [email protected]

ABSTRACT It is well known [1] that the natural geo-transmutation of chemical elements occurs in the atmosphere and in the regions of a strong change in geo-, bio-, acoustic-, and electromagnetic fields. The mineral row materials contain the same accompanying chemical combinations which are independent of mineral deposit [2]. This means that the formation of chemical elements occurs in the same physical and chemical conditions. These conditions were simulated on the fundamental cooperative resonance synchronization principle [1]. The experimental facility was constructed on the basis of our model which provided with the calculated final chemical elements. These experimental results indicate new possibilities for, simulating, inducing and controlling nuclear reactions by low energy external fields. The borrowing from the geo-transmutation mechanisms of chemical elements creates the fundamental directions in low energy nuclear reaction researches for construction of new ecologically pure mobile sources of energy independent of oil, gas and coal, new substances, and technologies.

References [1] F.A. Gareev, I.E. Zhidkova, E-print arXiv Nucl-th/0610002 2006. [2] V.A. Krivzskii, Transmutazija ximicheskix elementov v evolyuzii Semli (in Russian), Moscow 2003.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

POTENTIAL OF SOLAR HOME SYSTEMS IN PAKISTAN

Mujeebudin Memon8'*, Khanji Harijanb and Mohammad Aslam Uqaili0 a Professor, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan bPhD. Student, Department of Mechanical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan c Professor, Department of Electrical Engineering, Mehran University of Engineering and Technology, Jamshoro 76062, Pakistan E-mail: [email protected]

ABSTRACT About 68% of the population of Pakistan resides in rural areas. Most of the rural households have no access to electricity and meet lighting requirements through kerosene which is a major source of indoor air pollution and other environmental and health hazards. Rural villages are scattered over a large area and located far from the main electric grids. They have low population density and requires small load. About 67% of the conventional electricity in Pakistan is generated from fossil fuels with 51% and 16% share of gas and oil respectively. The indigenous reserves of oil and gas are limited and the country heavily depends on imported oil. The oil import bill is a serious strain on the country's economy. The combustion of fossil fuels also causes serious environmental pollution. The conventional power is even not sufficient for meeting the growing demand of electricity from the existing customers. Further more the extension of existing centralized grid system to far away from grid line rural areas with very low population density and small-scattered loads are economically and technically unfeasible. Hence there are remote chances of getting grid connection to most of the rural population in the near future. This whole situation requires urgent measures on priority basis for the development of indigenous, environment friendly, renewable energy sources such as solar energy.

This paper presents the assessment of potential of solar home systems (SHS) for rural electrification in Pakistan. The country lies in an excellent solar belt range and receives 16-21 MJ/m2 per day of solar radiation as an annual mean value, with 19 MJ/m2 per day over most areas of the country. It is estimated that about 7 million households in Pakistan do not have access to electricity (in 2004). Assuming that about 50% of the households in rural areas without electricity today would be electrified up to 2010, and only 25% of the remaining households could afford and would be willing to pay for a SHS; a potential market of about 875 thousand units would be realistic. Assuming a household would be equipped with an 80 Wp SHS which is sufficient for meeting the electricity needs, the respective capacity would amount to 70 MW. The technical potential of SHS. applications is estimated at 128 GWh/y. The electrification of rural households through SHS would minimize the environmental and health hazards and improves the socio-economic conditions of the rural people of Pakistan.

Keywords: Solar energy; Rural electrification; Environment friendly; Pakistan ^Corresponding author. Tel: +92-22-2772280; Fax: +92-22-2771382,

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

HEAT AND MASS TRANSFER ANALYSIS IN INTERMADIATE TEMPERATURE SOLID OXIDE FUEL CELLS (IT-SOFC)

Bora Timurkutluk, Mahmut D. Mat and Yüksel Kaplan Niğde Üniversitesi, Mühendislik Mimarlik Fakültesi Makine Müh. Bölümü, 51100 Niğde E-mail: mdmat@Niğde.edu.tr

ABSTRACT Solid oxide fuel cells (SOFCs) have been considered as next generation energy conversion system due to their high efficiency, clean and quite operation with fuel flexibility. To date, yittria stabilized zirconia (YSZ) electrolytes have been mainly used for SOFC applications at high temperatures around 1000°C because of their high ionic conductivity, chemical stability and good mechanical properties. However, such a high temperature is undesirable for fuel cell operations in the viewpoint of stability. Moreover, high operation temperature necessitates high cost interconnect and seal materials. Thus, the reduction in the operation temperature of SOFCs is one of the key issues in the aspects of the cost reduction and the long term operation without degradation as well as commercialization of the SOFC systems. With the reducing temperature, not only low cost stainless steels and glass materials can be used as interconnect and sealing materials respectively but the manufacturing technology will also extend. Therefore, the design of complex geometrical SOFC component will also be possible.

One way to reduce the operation temperature of SOFC is use of an alternative electrolyte material to YSZ showing acceptable properties at intermediate temperatures (600-800°C). As being one of IT-SOFC electrolyte materials, gadolinium doped ceria (GDC) has been taken great deals. In this study, a mathematical model for mass and heat transfer for a single cell GDC electrolyte SOFC system was developed and numerical solutions were evaluated. In order to verify the mathematical model, set of experiments were performed by taking species from four different samples randomly and five various temperature measurements. The numerical results reasonably agree with experimental data.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

NEW ENERGY OPINION LEADERS' LIFESTYLES AND MEDIA USAGE - APPLYING DATA MINING DECISION TREE ANALYSIS FOR UNIDO-ICHET WEBSITE USERS

Mavis Tsai, Ayfer Veziroğlu, Scott Warren, Yunze Que Department of Radio, Television and Film, Shih Hsin University #1, Lane 17, Mu-Cha Road, Sec. 1, Taipei, Taiwan(R.O.C) 116 office phone no: + 886(2) 22368225 ext. 3206 fax no:+ 886(2) 22361741 E-mail: [email protected]

ABSTRACT According to the innovation diffusion research, the innovators, opinion leaders, and diffusion agents play vital roles in promoting the acceptance of innovation. The innovators and opinion leaders must be able to cope with the high degree of uncertainty about an innovation and usually they have higher innovation-related media usage than the majority. Based on consumer behavior studies, lifestyle analysis could help researchers divide consumers into different lifestyle groups to understand and predict consumer behaviors. Lifestyle allows researchers to investigate consumers via their activities, interests and opinions instead of using demographic variables. The purpose of this research is to investigate how new energy innovators and opinion leaders' different lifestyles affect their new energy product adoption, and their media usage regarding new energy reports or promotion. In order to achieve the purposes listed above, the researchers need to locate and contact the potential innovators and opinion leaders in this field. Thus the researchers cooperate with UNIDO-ICHET to launch this survey. This cross-discipline online survey was formally launched from Aug 2005 to Oct 2006. The result of this survey successfully collected 2040 new energy innovators and opinion leaders' information. The researchers analyzed the data using SPSS statistics software and Data Mining decision tree analysis. Then the researchers divided new energy innovators into four groups: social-oriented, young modern, conservative, and show-off-oriented. They also analyzed which lifestyle groups are better targets for innovation agencies to launch innovation-related promotions or campaigns.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

ENERGY CONUNDRUM, DIGITAL REVOLUTION AND POLITICS

Ömer Ersun Ambassador (Retired), Türkiye E-mail: [email protected]

ABSTRACT The 21st Century will be fundamentally different from the previous one in all aspects of the human life. The world is now facing unprecedented challenges that will determine the fate of the human race as a whole. Our tiny planet is too small to shoulder the weight of six and a half billion energy-needy people and it is too vulnerable to afford violent and confrontational approaches as was the case in the past 20th century. It is also a fact that science opened new horizons before us. Digital Revolution inaugurated a new era in human history. Technology offers tremendous opportunities to overcome new and inherited problems. Sadly, the family of nations is ill-equipped for handling these challenges because the organizational structure of the world society is archaic and inoperative. Or, we live in a geostrategic environment pregnant with dangerous crisis of global significance. Furthermore, the good old days when scientists were heeded respectfully as reliable guides and when scientific facts were accepted as "veritas" are over. Solid scientific arguments are perceived as cover up stories to defend financial interests of multinational companies. Similarly, confidence in politicians is at its lowest level in several countries. At the center of this puzzle lies a frenetic quest for cleaner, cheaper and more secure energy sources.

In such circumstances, the best remedies which may be created by the brightest minds of the world will be tributary to the "goodwill" of politicians. Or, politicians are under the overwhelming pressure of their respective public opinions who may act according to emotional factors or advices from religion, gossip or ideology. Consequently, "Societal Issues" will be "the decisive" factor in shaping the future of "Emerging Nuclear Energy Systems" as well as non- nuclear technologies. This paper will attempt to identify major elements of this global equation from a political standpoint.

One also has to take into account a group of new and powerful actors: "The Big Emerging Markets". (China, India, Brazil, Mexico, South Korea, Türkiye, South Africa, Poland, Argentina, and Indonesia are "The Big Ten"). Together they are the most interesting consumers in the world energy market. Türkiye, as one of them, will be the subject of a brief case study in this paper. We will look at the options available to Türkiye, the fastest growing economy in the region, in making strategic choices among alternative energy sources.

Neither market forces alone, nor progress in new technologies will solve our problems. We need a new understanding, at the level of public opinions and the politicians, on the urgency and the gravity of the unique and particular conditions of the 21st century, and a holistic approach for tackling them efficiently.

212 PARALLEL SESSION 13A: FUSION II TR0700391

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

D-3HE FUELED FUSION DEVICES AS A STEP TOWARDS TOTAL FUSION SAFETY

Laila El-Guebaly1 and Massimo Zucchetti2 1 University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706, USA 2 Politecnico di Torino, Corso Duca degli Abruzzi 24 - 10129 Torino, Italy E-mail: [email protected]

ABSTRACT Nuclear fusion is seen as a "clean" source of energy. However, the attractive safety and environmental potential of fusion can only be fully realized by reducing the impact of materials activation and tritium inventory. The stress on fusion safety has stimulated worldwide research for fuel cycles other than D-T. With advanced cycles, such as D-3He, it is not necessary to breed and fuel large amounts of tritium. The D-3He fuel cycle in particular is not completely aneutronic due to the side reactions. Neutron wall loadings, however, can be kept low (by orders of magnitude) compared to D-T fueled plants with the same output power, eliminating the need for replacing the first wall and shielding components during the entire plant lifetime. Other attractive safety characteristics include low activity and decay heat levels, low-level waste, and low releasable radioactive inventory from credible accidents.

There is a growing international effort to alleviate the environmental impact of fusion and to support the most recent trend in radwaste management that suggests replacing the geological disposal option with more environmentally attractive scenarios, such as recycling and clearance [1,2,3]. We took the initiative to apply these approaches to existing D-3He designs: the ARIES-III power plant [4] and the Candor experiment [5]. Furthermore, a comparison between the radiological aspect of the D-3He and D-T fuel cycles was assessed and showed notable differences [6].

For the ARIES-III power plant, we estimated the highest possible activity to evaluate the disposal, recycling, and clearance options for managing the radwaste after decommissioning [6]. We compared the results to a D-T system to highlight the differences and the environmental impact. The results show that all ARIES-III in-vessel components qualify as Class A waste, the least hazardous type based on the U.S. guidelines. Potentially, all components can be recycled using conventional and advanced remote handling equipment. The bioshield contains traces of radioactivity and can be cleared from regulatory control after a relatively short period of time (-10 y).

Results obtained for the Candor experiment indicate that no environmental problems arise from such a device, from the radiological point of view, even with the presence of D-T plasma triggering. Candor does reach the zero-waste option as all wastes can be cleared within 100 y [6].

The D-3He cycle offers safety advantages and could be the ultimate response to the environmental requirements for future nuclear power plants. Furthermore, the low neutron production helps overcome some of the engineering and material hurdles to fusion development. Studies for the development of advanced fuel cycles should be carried out in parallel with the current mainstream fusion pathway that primarily focuses on D-T tokamaks, such as ITER, test facilities, DEMO, and power plants.

214 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

References 1. L. El-Guebaly, "Evaluation of Disposal, Recycling, and Clearance Scenarios for Managing ARIES Radwaste after Plant Decommissioning," 8th IAEA TM on Fusion Power Plant Safety (July 10-13, 2006, Vienna, Austria). To be published in Nuclear Fusion (2007). 2. M. Zucchetti, L. El-Guebaly, R. Forrest, T. Marshall, N. Taylor, and K. Tobita, "The Feasibility of Recycling and Clearance of Active Materials from a Fusion Power Plant," ICFRM-12, Dec. 4-9, 05, Santa Barbara, CA. To be published in Journal of Nuclear Materials (2007). 3. L. El-Guebaly, R. Pampin, and M. Zucchetti, "Clearance Considerations for Slightly- Irradiated Components of Fusion Power Plants," 8th IAEA TM on Fusion Power Plant Safety (July 10-13, 2006, Vienna, Austria). To be published in Nuclear Fusion (2007). 4. F. Najmabadi, R. Conn et al., "The ARIES-III D-3He Tokamak-Reactor Study," Proceedings of IEEE 14th Symposium on Fusion Engineering, San Diego, CA, Vol. 1,213 (Sept. 30-Oct. 4,1991). 5. B. Coppi, P. Detragiache, S. Migliuolo et al., "D-3He Burning, Second Stability Regionand the Ignitor Experiment," Fusion Technology, Vol. 25, 353 (1994). 6. L. El-Guebaly and M. Zucchetti, "Recent Developments in Environmental Aspects of D- 3He Fuelled Fusion Devices," to be published in Fusion Engineering and Design.

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RECONSTRUCTION AND ANALYSIS OF TEMPERATURE AND DENSITY SPATIAL PROFILES FROM INERTIAL CONFINEMENT FUSION IMPLOSION CORES

R. C. Mancini Physics Department, University of , Reno, USA E-mail: [email protected]

ABSTRACT We discuss several methods for the extraction of temperature and density spatial profiles in inertial confinement fusion implosion cores based on the analysis of the x-ray emission from spectroscopic tracers added to the deuterium fuel. The ideas rely on (1) detailed spectral models that take into account collisional-radiative atomic kinetics, Stark broadened line shapes, and radiation transport calculations, (2) the availability of narrow-band, gated pinhole and slit x-ray images, and space-resolved line spectra of the core, and (3) several data analysis and reconstruction methods that include a multi-objective search and optimization technique based on a novel application of Pareto genetic algorithms to plasma spectroscopy. The spectroscopic analysis yields the spatial profiles of temperature and density in the core at the collapse of the implosion, and also the extent of shell material mixing into the core. Results are illustrated with data recorded in implosion experiments driven by the OMEGA and Z facilities.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

STUDY OF CONVERSION EFFICIENCIES FOR LASER LIKE A VISTA IN THE LASER PLASMA EXPERIMENTS AND PIC-SIMULATIONS

Yu. P. Zakharov, K. V. Vchivkov, A. V. Melekhov, V. G. Posukh, E. L. Boyarintsev, I. F. Shaikhislamov, H. Nakashimaa Institute of Laser Physics (ILP), Novosibirsk, 630090, Russia E-mail: [email protected] a Kyushu University (KU), 6-1 Kasuga-Koen, 816-8580, Japan

ABSTRACT Space Power and Propulsion Last years a lot of numerical simulations by 3D/PIC-code of KU [1,2] and first laboratory simulative experiment of ILP [3] with Laser-produced Plasma (LP) in dipole {\x) magnetic field

had proved that a real conversion efficiency r\9 (of plasma ions momentums) could reach enough

high value r\p > 50% in a VISTA-scheme [4]. While the another important item of VISTA design

- to achieve at least a 5%-level of direct conversion efficiency r\e (of plasma energy into pick-up coils) is still a difficult problem under conditions of possible strong influence of various instabilities of plasma onto its conductivity and diamagnetism. First experimental data on a very

low level of r\e ~ 0.5% were obtained in the MHD experiment [5] under geometry (of LP-

explosion at axial distance X ~ Rt from field-coil of this radius) rather close to VISTA-design, but under relative energetic conditions very far from it. For initial plasma energy Eo, a relevant energetic parameter ee ~ 3EOX3/JJ,2 ~ 0.01 for MHD case, but for VISTA could have a finite value ee ~ 0.3 close to its critical value 0.4, above which plasma should escape field. So a VISTA-KI experiment is planned now under physical and geometrical conditions well corresponding to

VISTA case: a=X/Rt «1, se > 0.1 and enough low value [6] of ion magnetization criterion Sb < 0.5. These conditions will be realized by using: a spherically expanding LP (from small pellet irradiated by 4 beams) with Eo ~ 20 J and front velocity Vo ~ 150 km/s, field-coils system with 3 effective \i = 2-^5 MG*cm and Rt « 7 cm, combined with the same size pick-up coil (of inductance L) and its Rogowski coil to measure induced current J. To predict a plasma and magnetic field dynamics in VISTA-KI experiment we did its numerical simulations by KU-code that was tested by the data of MHD experiment with uniform field Bo « 8 kG. Main result of this testing shows that in the runs with grid size 1.5 mm, a maximal diamagnetic cavity radius achieves theoretical one Rb = (3Eo/Bo2)1/3 «2 cm and the corresponding changing of magnetic flux AO at axial distance 4.2 cm (in a plane of one field-coil with Nt= 15 of turns, acting there as one-half of pick-up coil) gives us expected ideal current value Jm(A) = A was smaller in a good agreement with the measured current 33 A in MHD. Thus we can use this simplified expression for Jm to estimate the 2 efficiency r\e °c Jm /LEo, but due to effects of real plasma we should use only one-half of its ideal value, denoted as *r\e. First results of VISTA-KI numerical simulation confirm that under the same assumptions we can use such Jm-estimation of r\e in a quasi-dipole field also (at least for 3 ae<0.1) and therefore we can rewrite it into general form [7] as *r\s(%) « (5*10)ae /a from which we can conclude that for parameters of VISTA its efficiency *r\e < 2% could be very low. So the

217 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye main goal of VISTA-KI experiment and its further numerical simulations will be verification of proposed r\e (as,a) relation under conditions of t|p > 50% (and 8b < 1), to optimize original VISTA-scheme [4].

* This work was supported by Russian Fund of Basic Research, Grant 05-08-50068. References 1. Nagamine Y. and Nakashima H., Fusion Technology, 35 (1999) 62. 2. Vchivkov K., Nakashima H., Zakharov Yu.P., et al, Jpn. J. Appl Phys., 42 (2003) 6590. 3. Zakharov Yu.P., et al., in Proc. 4th Symp. "Current Trends in Intern. Fusion Research" (Wash., DC, 2001) 31. 4. Orth, CD., LLNL Report VISTA, University of California: UCRL-LR-110500 (2003) 143 p. 5. Zakharov Yu.P., et al. Trans. Fusion Science and Technology, 41 (2005) 187. 6. Zakharov Yu.P., et al., Plasma Physics Reports, 32 (2006) 183. 7. Zakharov Yu. P., In "Inertial Fusion Sciences and Applications: State of the Art 2003" (American Nuclear Society, 2004) 1106.

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13* International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ON THE POSSIBILITY OF D-3HE FUSION BASED ON FAST-IGNITION INERTIAL CONFINEMENT SCHEME

Yasuyuki Nakao1, K. Hegi, ! T. Ohmura1, M. Katsube1, T. Johzaki2, K. Kudo1 and M. Ohta3 1 Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395, Japan 2 Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871, Japan 3 Teikyo University Fukuoka College, 4-3-124 Shinkatsutachicho, Omuta, Fukuoka 836- 8505,Japan E-mail: [email protected]

ABSTRACT Although nuclear fusion reactors adopting D3He fuel could provide many advantages, such as low neutron generation and efficient conversion of output fusion energy, the achievement of ignition is a difficult problem. It is therefore of particular importance to find some methods or schemes that relax the ignition requirements. In inertial confinement scheme, the use of pure D3He fuel is impractical because of the excessive requirement on driver energy. A small amount of DT fuel as "igniter" is hence indispensable.1 Our previous burn simulation1 for DT/D3He fuels compressed to 5000 times the liquid density showed that substantial fuel gains (-500) are 2 2 obtained from fuels having parameters p RDT = 3 g/cm , p R iotai = 14 g/cm and a central spark temperature of 5 keV. The driver energy needed to achieve these gains is estimated to be~30 MJ when the coupling efficiency is 10%; in this case the target gain is -50. Subsequent implosion simulation2, however, showed that after void closure the central DT fuel is ignited while the bulk of the main D3He fuel is still imploding with high velocities. This pre-ignition of DT fuel leads to a low compression of the main fuel and prevents the DT/D3He fuel from obtaining required gain. These difficulties associated with the pre-ignition of DT fuel could be resolved or mitigated if other ignition schemes such as fast-ignition3 and/or impact-ignition4 are adopted, because in these schemes compression and ignition phases are separated. Furthermore, the reduction of driver energy can be expected. In the present study, we examine the possibility of D3He fusion in the fast-ignition scheme. Simulations until now have been made for a DT/D He fuel compressed to 5000 times the liquid density by using FIBMET (2D fusion ignition and burning code) 5 and a newly developed neutron diffusion code. DT igniter was assumed to be placed at a corner of the compressed fuel. The p R values and temperature of compressed fuel were assumed as p RDT = 4 g/cm2, p R total = 12 g/cm2 and 0.2 keV. The coupling efficiencies of implosion and heating lasers were respectively taken as 10% and 30%. The work shows that it is possible to obtain sufficient target gains (-60) with realistic driver energy below 10 MJ (-8 MJ for implosion plus -0.3 MJ for heating). Crucial role of DT fusion neutrons in the D3He main fuel heating was clarified. The possibility to reduce the amount of DT igniter will be discussed.

References 1. T. Honda, Y. Nakao, Y. Hnada, K. Kudo, H. Nakashima, Nucl. Fusion, 31, 851 (1991); Y. Nakao, T. Honda, H. Nakashima, Y. Honda, K. Kudo, Fusion Technol., 20, 66 (1992). 2. H. Nakashima, M. Shinohara, Y. Wakuta, T. Honda, Y. Nakao, H. Takabe, Laser Part. Beams, 11,137(1993).

219 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

3. M. Tabak, J. Hammer, M.E. Glinsky, W.L. Kruer, S.C. Wilks, J. Woodworm, E.M. Campbell, M.D. Perry, R.J. Mason, Phys. Plasmas, 1, 1626 (1944). 4. M. Murakami, H. Nagatomo, H. Azechi, F. Ogando, M. Perlado, S. Eliezer, Nucl. Fusion, 46, 99 (2006) 5. T. Johzaki, K. Mima, Y. Nakao, H. Nagatomo, A. Sunâhara, Proc. of 3rd Int. Conf. on Inertial Fusion Sciences and Applications, Monterey, 2003, edited by B.A. Hammel, et al. (LLNL, 2004), p. 474.

220 PARALLEL SESSION 13B FISSION REACTORS VI TR0700395

13lh International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

INVESTIGATION OF THE PROPERTIES OF THE NUCLEI USING ON THE NEW GENERETION REACTOR TECNOLOGY SYSTEMS

E. Tel1, H. M. Şahin2, Şenay Yalçın3, Taner Altınok4, A. Kaplan5' A. Aydın6, 1 Gazi University, Faculty of Arts and Science, Department of Physics, Beşevler, ANKARA- TÜRKİYE 2 Gazi University, Faculty of Technical Education, Department of Mechanical Education, ANKARA- TÜRKİYE 3Bahçeşehir Üniversitesi, Mühendislik Fakültesi, Beşiktaş, İstanbul, Türkiye 4Kara Harb Okulu, Savunma Bilimleri Enstitüsü, Ankara 06654, Türkiye 5 Süleyman Demirel University, Science and Letter Faculty, Department of Physics, ISPARTA- TÜRKİYE 6Kınkkale University, Faculty of Art and Science, Kırıkkale, TÜRKİYE E-mail:

ABSTRACT The application fields of the fast neutron are Accelerator-Driven subcritical Systems (ADS) for fission energy production and hybrid reactor systems. The technical design hybrid reactor and ADS systems potentialities require the knowledge of a wide range of better data and much effort. Thorium (Th) and Uranium (U) are nuclear fuels in these reactor systems. Lead (Pb), Bismuth (Bi) and Tungsten (W) are the target nuclei in the ADS reactor systems. The Hartree-Fock (H-F) method with an effective interaction with Skyrme forces is widely used for studying the properties of nuclei such as binding energy, Root Mean Square (RMS) charge radii, mass radii, neutron density, proton density, electromagnetic multipole moments, etc. In this study, by using H-F method with interaction Skyrme RMS charge radii, RMS mass radii, neutron density and proton density were calculated for the 232Th, 238U, 207Pb, 209Bi and mW isotopes used on the new generation reactor systems. The calculation results of charge radii have been compared with experimental data and obtained other results have been discussed for hybrid and ADS reactor systems.

Keywords: Fertile- fissile materials, Skyrme force, Neutron and proton density

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

REACTIVITY INSERTION ACCIDENT IN A SMALL MOLTEN SALT REACTOR

Yoichiro Shimazu*, Nobuhide Suzuki** *Graduate School of Engineering, Hokkaido University, Japan ** Kobe Shipyard, Mitsubishi Heavy Industries, Inc. Tel. and fax. +81-11-706-6676

E-mail; shimazu@eng. hokudai. ac.jp

ABSTRACT Molten Salt Reactors (MSRs) have a long history with the first design studies beginning in the 1950's at the Oak Ridge National Laboratory (ORNL). MSRs have many advantages such as improved potential in safety, proliferation resistance, resource sustainability and waste reduction. But MSR developmental activities have slowed considerably except for some small scale efforts, mostly in Russia, France, Japan and a few other places and there are few data available to support detailed analyses. Recently, a conceptual design of a small MSR, name Fuji-12 has been proposed. Fuji-12 operates with the same fuel salt as the Molten Salt Breeder Reactor (MSBR) designed by ORNL. But it differs from the ORNL design in several ways, such as no on-site chemical processing plant and a low rated power. The authors are interested in the MSR concept due to its high potential in the areas of safety, proliferation resistance, resource sustainability and waste reduction, all necessary requirements for the generation IV nuclear power systems. Therefore the MSR concept has been selected as one of the more promising candidates for future consideration.

The authors believe that additional investigations are necessary for future study. From this point of view, the authors have analyzed various reactivity insertion accidents due to control rod malfunctions in a MSR named FUJI-12. The MSR can be operated with a small excess reactivity and also the control rods for power adjustment consist of graphite, which has buoyancy in the fuel salt. Thus the reactivity addition could be limited by design. However at the same time the delayed neutron fraction is quite small due to the usage of U-233 as fissile material and the circulation of the fuel salt. Therefore reactivity insertion accident should be qualitatively evaluated. The reactor transients were analyzed without in order to evaluate the severity of such accidents against the safety.

The severity of the accident was discussed for the outlet fuel temperature in comparison with the limiting temperature of the reactor vessel integrity. Although the total primary system design of FUJI-12 is not completed and thus the accident analysis includes some crude assumptions, it can be expected that the reactivity insertion accident in FUJI-12 would not result in a severe plant conditions.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

COMPUTATIONAL ANALYSIS OF BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM

J. Seok Hwang, Hyoung M. Son, W. Soo Jeong, Tae W. Kim, Kune Y. Suh

Seoul National University, San 56-1 Sillim-dong, Gwanak-gu, Seoul, 151-744, Korea E-mail: [email protected]

ABSTRACT Battery Optimized Reactor Integral System (BORIS) is being developed as a multi-purpose fast- spectrum reactor cooled by lead (Pb). BORIS is an integral optimized reactor with an ultra-long- life core. BORIS aims to satisfy various energy demands maintaining inherent safety with the primary coolant Pb, and improving economics. BORIS is being designed to generate 23 MWj, with 10 MWe for at least twenty consecutive years without refueling and to meet the Generation IV Nuclear Energy System goals of sustainability, safety, reliability, and economics. BORIS is conceptualized to be used as the main power and heat source for remote areas and barren lands, and also considered to be deployed for desalinization purpose. BORIS, based on modular components to be viable for rapid construction and easy maintenance, adopts an integrated heat exchanger system operated by natural circulation of Pb without pumps to realize a small sized reactor. The BORIS primary system is designed through an optimization study. Thermal hydraulic characteristics during a reactor steady state with heat source and sink by core and heat exchanger, respectively, have been carried out by utilizing a computational fluid dynamics code and hand calculations based on first principles. This paper analyzes a transient condition of the BORIS primary system. The Pb coolant was selected for its lower chemical activity with air or water than sodium (Na) and good thermal characteristics. The reactor transient conditions such as core blockage, heat exchanger failure, and loss of heat sink, were selected for this study. Blockage in the core or its inlet structure causes localized flow starvation in one or several fuel assemblies. The coolant loop blockages cause a more or less uniform flow reduction across the core, which may trigger coolant temperature transient. General conservation equations were applied to model the primary system transients. Numerical approaches were adopted to discretize the governing equations. A computer program was developed based on the energy and momentum equations. The fluid velocity and properties variation were averaged over the cross sectional area of the flow path. The Boussinesq approximation was used to simplify the momentum equation. Latest friction factor correlations were supplied to accurately simulate the coolant circulation of the system. The finite element approach was adopted to calculate the fuel element temperature distribution to see whether specified design limit is conserved. Analyses indicate the availability of acceptable margins against the design safety limits in all the parametric cases analyzed.

224 TR0700398

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PRELIMINARY STUDY ON CHARACTERISTICS OF EQUILIBRIUM THORIUM FUEL CYCLE OF BWR

Abdul Waris1'3, Rizal Kurniadi1, Zaki Su'ud1, and Sidik Permana2

department of Physics, Bandung Institute of Technology Jl. Ganesa 10 Bandung, 40132, Indonesia ^Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550, Japan 3E-mail: awaris@fi. itb. ac. id

ABSTRACT One of the main objectives behind the transuranium recycling ideas is not merely to utilize natural resource that is uranium much more efficiently, but to reduce the environmental impact of the radio-toxicity of the nuclear spent fuel. Beside uranium resource, there is thorium which has three times abundance compared to that of uranium which can be utilized as nuclear fuel. On top of that thorium is believed to have less radio-toxicity of spent fuel since its produce smaller amount of higher actinides compared to that of uranium. However, the studies on the thorium utilization in nuclear reactor in particular in light water reactors (LWR) are not performed intensively yet. Therefore, the aim of the present study is to evaluate the characteristics of thorium fuel cycle in LWR, especially boiling water reactor (BWR). To conduct the comprehensive investigations we have employed the equilibrium burnup model 1"3). The equilibrium burnup model is an alternative powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor^. We have employed 1368 nuclides in the equilibrium burnup calculation where 129 of them are heavy metals (HMs). This burnup code then is coupled with SRAC cell calculation code by using PIJ module to compose an equilibrium-cell burnup code. For cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been used in this study. The design parameters of the studied BWR are presented in Table 1, which is the design parameter of the BWR/6 of the General Electric's BWR. Keywords: thorium fuel cycle, equilibrium burnup, BWR

Table 1 Design parameters of studied BWR Power Output (Thermal) 3000 MW Average power density 50 Wcm-3 Radius of fuel pellet 0.529 cm Radius of Fuel rod 0.615 cm Pin Pitch 1.444 cm Void coefficient 42% Fuel type Oxide Cladding Zircaloy-2 Coolant H2O

225 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

References 1. A. Waris and H. Sekimoto, "Characteristics of several equilibrium fuel cycles of PWR", J. Nucl. Sci. Technol, 38, p.517-526, 2001 2. A. Waris, H. Sekimoto, and G. Kastchiev, Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycles of PWR, Int. Conf. On the New Frontiers of Nuclear Technology, PHYSOR 2002, Seoul, Korea, 2002 3. A. Waris, et. al., Influence of void coefficient change on Plutonium and Minor Actinides Recycling in BWR with Equilibrium Burnup, Proc. Int. Conf. Innovative Nuclear Energy Systems, INES2 2006, November 26-30, 2006, Yokohama, Japan, 2006.

226 PARALLEL SESSION 13C: SOCIETAL ISSUES TR0700399

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

TECHNICAL CONSIDERATIONS IN NUCLEAR TERRORISM

Professor Guillermo Velarde President Institute of Nuclear Fusion, Jose Gutierrez Abascal n° 2 Madrid 28006 - Spain E-mail: gvelarde(a),denim.upm.es

ABSTRACT Nuclear terrorism is an evil application of nuclear energy, in the same way that chemical and biological terrorism could be considered as the evil side of chemistry and biology. This paper presents two effects of nuclear terrorism. First, dirty bombs or radioactive bombs or radiological dispersion devices (RDDs), and second, crude atom bombs or improvised nuclear devices (INDs). The paper analyses as well the probabilities of an attack, its biological effects and nuclear risk.

Experiments carried out so far indicate that the lethal effects produced by RDDs are likely the same that the effects produced by the chemical explosive used in the bomb. These type of bombs are rather bounded to generate panic and have implicit a high cost of decontamination. It will be described the measures to be adopted.

INDs will be also considered. Uranium INDs by gun-method are more feasible to be made. They can be disassembled and their components transported to the target place. Plutonium INDs by the implosion-method are complex and required high precision technology. Their disassembly is very difficult.

This paper analyzes too the illicit aspect of uranium and plutonium traffic.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

SOCIETY RESPONSE TO NUCLEAR ENERGY

Professor Natividad Carpintero Santamaria General Secretary, Institute of Nuclear Fusion - Polytechnic University of Madrid C/ Jose Gutierrez Abascal n° 2 - 28006 Madrid (Spain) E-mail: ncsantamaria@telefonica. net

ABSTRACT Energy demand in the world is growing increasingly, among other factors due to economic development. Every way of producing electricity has got their own drawbacks and has implicit environmental impact. Among all the energy sources, nuclear energy is the most polemic because of the way it is presented by the mass media. This aspect provokes controversy to occidental societies which reject this kind of energy with arguments normally based on a wrong and insufficient knowledge of the matter. The antinuclear discourse, promoted late in the seventies, has gone deeply into the collective social unconscious and has undermined public acceptance of nuclear energy due to the fact, deeply exploited by antinuclear groups, of linking nuclear energy with the atomic bombing of Hiroshima and Nagasaki. In this sense, it is important to mention that in Japan there was a profound resentment and opposition to nuclear energy, because the memory of the nuclear bombings was permanently alive. However when the Japanese government told its people that this energy was necessary to boost their industrial development, Japanese citizens in an unprecedented attitude of patriotism overcame their most antagonist feelings, in order to contribute to the industrial development of their country. The result was that most of them voted in favour. Presently Japan gets 30% of its energy by means of 56 nuclear power plants and 1 more is under construction. Antinuclear groups took as their best emblem the accident of Chernobyl to justify their opposition to the nuclear power plants. The manipulation of this accident has been one of the most shameful in the nuclear history. It is widely known among the experts that the reactor used in Chernobyl was a type of military plutonium converter with a positive temperature reactivity coefficient, which made very dangerous its functioning. Any nuclear regulatory commission in democratic and responsible countries would have never authorized the use of this reactor. For this reason, it is very important to tell the truth to public opinion leaving out tendentious and demagogic positions. The energetic future will be based on nuclear fusion, since one of its fuels is deuterium, found in the water and accessible to any country in the world. Nuclear fusion is the best alternative to the present dependence of primary energies based on fossil fuel sources. This paper presents different alternatives to improve the image of nuclear energy among population in a moment in which several countries present an onset of crisis in the energetic sector due to the economic growth. The fact is that the demand may start to be overcome by the offer and this circumstance, together with the global warming of our planet as reported by the United Nations, have provoked a new perspective in the debate of boosting nuclear energy as a fundamental source because it is ecologically and economically sound.

229 TR0700401

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

USEFULLNESS OF UNDER CONTROLLED POWERS

M. Oktay ALNIAK Bahçeşehir University, Faculty of Engineering Istanbul - TÜRKİYE E-mail: oalniak@bahcesehir. edu. tr

ABSTRACT In the view of production of the energy with the aspect of healthy, confidential, sustainable and profitable; severity of the nuclear energy directly realized rather than alternative energies. There are a lot of studies holds on the purpose of preferring to use pure energy resources rather fossil resources of patrol to satisfy the energy demand. Hydrogen usage at the transportation, heating and energy applications should prefer than fossil fuels. Also production of the hydrogen from a nuclear power resource by the cost-effective way is important to the project of a clean world. Türkiye should also benefit from that energy among 30 countries in the world take its advantages. On the other hand putting on the agenda of bad experiences at the nuclear energy applications, technical problems, which are possible for all another applications, should not be barrier improvements at that area. We can not warm up in the Türkiye! We have not got enough money. We feel cold... Human being misses to be accustomed life standards. For this reason keeping away from nuclear energy's opportunity looks like keeping away from civilization. That is accepting live in dark ages. This energy's safety usage requires education and technology Production of this energy brings potential and strategic power to the country. Power should produced and used if it is controlled. First controlling way is having well educated brains, scientific methodology and technology that is satisfy this control. After that producing energy appropriate process with control. Passing to the charcoal from wood heating from charcoal to patrol and stop there, is likely a case as failing as at the civilization class. Countries could become civilized as they worked and deserved. Türkiye is in a geopolitical and geostrategical geography between the Balkans and Middle East. She's also a bridge between Europe and Asia regarding exchange of cultures, civilization and industrialization. As an important developing country in the region Türkiye needs sustainable energy for her development. Nuclear energy facilities can only meet the requirements of Turkish growing rate in industrial field. When making strategic decisions for country's current situation and future, some main topics, as importance of necessity for energy and power, more time, work, education, money and organization are essential to decide. As a result the underlined factors will be improved as functional of function of functions.

230 TR0700402

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

IMPACT OF THE DECOMMISSIONING OF NUCLEAR FACILITIES AND RADIOACTIVE WASTE TRAFFICKING IN AFRICA Baba Gana Abubakar Nigeria Ports Authority, Senior Officer, Alhaji Bukar Kuya House, Oppossite Aburos Mosque, Fezzan, Maiduguri, Nigeria E-mail: babaganabubakar2002 @yahoo.com

ABSTRACT INTRODUCTION Africa is the world's second largest and the most populated continent after Asia, it has a total population of approximately 800 million people. It comprises of 54 sovereign nations out of which 36 are coastal countries and blessed with over 100 Seaports.

Apart from Nigeria, South Africa, Egypt, Libya, Morocco, Tunisia and Libya, all the other remaining African countries are extremely poor and unviable. As a result of this, Africa has been experiencing a lot of civil unrest since the 1960s when most of the African countries gained their independence from their former colonial masters, the civil unrest in countries like Angola, Democratic Republic of Congo, Sudan, Burundi, Rwanda, Mozambique, Liberia, Sierra Leon and recently in Cote D'lvoire, are good examples. In addition to abject poverty of less than 1$ per person per day makes trafficking in drugs, arms, humans and weaponry trade on the continent becomes much more rampant.

Today the continent is experiencing the coming of a new evil deal called "Trade in radioactive waste"; which involves the transporting of materials from existing or decommissioned nuclear plants ranging from fairly used Trucks, laboratory equipments, office facilities, clothing materials like booths and raincoats, roofing sheets and even toxic waste from the developed countries to it's waste bin in Africa, where it is unsafely disposed after collecting millions of dollars from It's original owners (UN report, 2001). Recent statistics have revealed that most of the people involved in the evil businesses of trafficking in drugs, human, arms and trading in weaponry, are diverting in to the so called new evil business of "Trade in Radioactive waste" because this new evil business financially exceeds the rest of the übove listed evil businesses. This is clearly proved by the recent toxic waste disposed in Abidjan Cot Devoir in August 2006.

The materials from the decommissioned nuclear plant sites can be hazardous if for example a roofing sheet that was once used in the site is brought to a town or a city and subsequently used in roofing other structures, because such roofing sheets may contain radioactive airborne particles there is a tendency that such particles can be washed away by rain and through run-off the rain water may reach the river, well, dam and even the ground water through "vicious circle" this situation may lead to the consumption of such particles and subsequently causes Cancerous related sicknesses in the community. Apart from that all the personnel in the various Sea Ports, Airports and Railways were such materials passed can be at risk of getting cancerous sicknesses.

231 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

POSITION OF THE AFRICAN GOVERNMENTS ON THE WASTE TRFFICKING. Apart from the government of the federal republic of Nigeria, that has made great fuss about wastes disposal and importation in to its land and further contributed to the issuance of a resolution by the African Union (AU). All the other governments of the other African countries have shown no sings of fighting this evil business in their lands, some governments were silent on the issue and some United Nations reports indicated that some others even sign treaties, contracts or received offers to approve the importation of wastes in their lands.

In addition there were also many other unsuccessful dumping of these wastes in Africa due to government efforts, examples of these are that of the koko port, were the personnel of the Nigerian Ports Authority (NPA), helped in stopping the importation of such wastes in 1988 where an Italian company was forced to evacuate its wastes earlier dumped at the koko port, in the present day Delta state of Nigeria. The hazardous waste exposed by the Tsunami disaster in December 2004 at the coast of Somalia also shows how vulnerable African coastal countries are to nuclear related wastes.

Since the history of waste trafficking in Africa has shown that the gateway of the importation is mainly the Sea Ports, this makes the African coastal Countries becoming more vulnerable to be used by the undesirables for the importation of the decommissioned sites facilities ranging from laboratory equipments, office facilities, Trucks, tyre and even clothing facilities like boots and rain coats among many others in the name of fairly used items.

POSITION OF NON-GOVERNMENTAL ORGANISATIONS (NGOS) ON WASTES TRAFFICKING As a matter of fact apart from Nigeria that is fighting this evil business both directly and indirectly, the NGOs are the only remaining bodies that are helping towards the fighting of this evil business and on this part of the world. For example the International Parliamentary Union helped in organizing and supervising a conference titled "Health is the Basis of Development in Africa" which was held in the Congolese capital, Brazzaville in 1990 where the issue of industrial toxic wastes disposal in Africa, was debated at a highly academic standard. Another conference was also organized by the NGOs on "Nuclear Pollution" in the Ghanaian Capital, Accra, in 1993. In fact the NGO's are trying, but more is still expected from them in order to educate people on the safe handling of the radioactive waste especially at disposal and to conquer this evil business of 'Trade in radioactive waste' moving with a supersonic speed in this part of the world.

POTENTIAL SOURCES OF NUCLEAR AND RADIOACTIVE WASTES IN AFRICA. The African continent is underdeveloped and also considered to be unviable. So there is no significant generation of Nuclear or the radioactive wastes from the industries. However, the existence of Uranium Mines in countries like Niger Republique and the Democratic republic of Congo proves that the continent also generates some radioactive wastes to some level. Niger is the fourth largest Uranium producer in the world; it produces 11% of the total world supply from mines while the Democratic Republic of Congo is the seventh largest producer with 7% of the total World supply.

232 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

Apart from the natural sources of generating radioactivity, the presence of nuclear plants in South Africa for the generation of electricity is also a potential source of generating nuclear decommissioned site and radioactive wastes. Other industrial sources include the pharmaceutical and the petroleum companies operating in countries like Egypt, Libya, Zimbabwe, South Africa, Morocco and some few others. THE INHABITANTS OF AFRICA AND THE NUCLEAR WASTES TRAFFICKING Recent research indicates that about 81% of the inhabitants of the African continent do not know anything about nuclear or radioactive wastes. In fact even the inhabitants living around the Uranium mines of "Arlit" which is one of the world's major uranium mines, in the Niger Republique, do not know what Uranium is, not to talk about its radioactivity.

As a result of this ignorance about radioactivity by the people of this Continent coupled with the rapidly growing trafficking and unsafe handling of this radioactive wastes by some few individuals, led to the out-break and the subsequence spreading of so many types of Cancerous tumours, the contamination of some farm lands rendering them uncultivable and the extinction of so many types of sea animals and sea foods e.g. The rapid decline of sea foods around the Somalia's territorial waters, the Gulf of Guinea and the situation in the Dollos Islands along the coast of Guinea-Bissau near the capital Bissau, are all good examples.

RECOMMENDATION After identifying the presence of this evil business of radioactive waste trafficking on the African continent, I came up with the following suggestions/recommendations: -

1. The International Atomic Energy Agency (IAEA), should partner with the Nigerian government and the Nigerian Ports Authority in particular in organizing workshops from time to time towards educating the personnel of the African Sea ports and Port Authorities in fighting this man made menace of importation of nuclear wastes from any part of the World and at the same time prevent future possible decommissioned sites materials.

2. There should be incessant, adequate and mass public enlightenment on the dangers of effects and the unsafe handling of radioactive materials (waste), by both the International Atomic Energy Agency (IAEA) in collaboration with the Governments of Africa and the Non governmental Organizations through sponsoring and organizing conferences and seminars from time to time.

3. The International Atomic Energy Agency should use its capacity to be able to influence the African Union (AU) and other African regional bodies to pass a resolution banning this evil business of trafficking of nuclear or radioactive wastes in this part of the World.

4. The International Atomic Energy Agency should send team of researchers to come and investigate this trend of radioactivity within the African continent and proffer possible lasting solutions in checking the menace. 5. The International Atomic Energy Agency, should be making a closer monitoring of the activities of the nuclear plants existing in like South Africa in order to avoid the problem of illegal dumping of their nuclear or radioactive wastes.

233 13 International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

ANALYSIS AND CONCLUSION We cited at the importation, subsequence and illegal traffic of nuclear related wastes in to the African coastal countries, in relation to the role of the individuals, the governments and the non- governmental organizations. Although because of the influence of high rate of poverty coupled with the underdevelopment of the continent resulted to the high traffic and illegal disposal of nuclear related wastes in the continent, but yet the Nigerian government and the Non- governmental Organizations (NGO) are trying on their own capacity towards combating this evil business.

Despite the single handed effort by Nigeria and the educative conferences organized by the NGO's towards combating this menace, the presence and the increasing cases of Cancerous tumours, the continuous contamination of cultivable lands, the rapid decline in sea foods and the continuous exposures of hazardous waste as a result of natural disasters like earthquakes and floods in the continent, shows that the problems emanating from the trafficking of nuclear and radioactive wastes keeps on increasing.

It was in view of these increasing problems, I came up with the above listed suggestions/recommendations with the hope that if these suggestions/recommendations are implemented and adopted, it will help in reducing health hazards and loss of lives through the trafficking of these nuclear and radioactive wastes in this part of the World, other wise the problem will ever remain on the increase.

234 PARALLEL SESSION 14A: EXOTIC CONCEPTS IN FUSION TR0700403

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

INERTIAL CONFINEMENT FUSION AND RELATED TOPICS

Alexander N. Starodub N.G Basov Institute of Quantum Radiophysics P.N.Lebedev Physical Institute ofRAS, Moscow, Russia E-m&il:sautkin@sci. lebedev. ru

ABSTRACT The current state of different approaches (laser fusion, light and heavy ions, electron beam) to the realization of inertial confinement fusion is considered. From comparative analysis a conclusion is made that from the viewpoint of physics, technology, safety, and economics the most realistic way to future energetics is an electric power plant based on a hybrid fission-fusion reactor which consists of an external source of neutrons (based on laser fusion) and a subcritical two-cascade nuclear blanket, which yields the energy under the action of 14 MeV neutrons. The main topics on inertial confinement fusion such as the energy driver, the interaction between plasmas and driver beam, the target design are discussed. New concept of creation of a laser driver for IFE based on generation and amplification of radiation with controllable coherence is reported. The performed studies demonstrate that the laser based on generation and amplification of radiation with controllable coherence (CCR laser) has a number of advantages as compared to conventional schemes of lasers. The carried out experiments have shown a possibility of suppression of small-scale self-focusing, formation of laser radiation pulses with required characteristics, simplification of an optical scheme of the laser, good matching of laser-target system and achievement of homogeneous irradiation and high output laser energy density without using traditional correcting systems (phase plates, adaptive optics, space filters etc.). The results of the latest experiments to reach ultimate energy characteristics of the developed laser system are also reported.

Recent results from the experiments aimed at studying of the physical processes in targets under illumination by the laser with controllable coherence of radiation are presented and discussed, especially such important laser-matter interaction phenomena as absorption and scattering of the laser radiation, the laser radiation harmonic generation, X-ray generation, crater formation and plasma expansion under laser pulse action, conversion of laser radiation by means of nonlinear crystals, influence of coherence degree on the processes mentioned above.

Corresponding author: A.N.Starodub, P.N.Lebedev Physical Institute of the RAS, Leninsky prospect, 53, Moscow, 119991, Russia, tel. 7(495)1350350, fax 7(495)9382251, e-mail: sautkin@sci. lebedev. ru,

236 TR0700404

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

STUDIES ON THE KINETICS OF MUON CATALYZED FUSION IN THE HT MIXTURE WITH VERY LOW TRITIUM CONCENTRATION

S. N. Hosseini Motlagh Tehran -Iran University Science & Technology E-mail: [email protected]. ir

ABSTRACT The idea of muon catalyzed fusion (/JCF )was first suggested by C.Frank in 1947 ,[l] when he tried to explain tracks from cosmic rays in photoemulusions exposed at high altitudes .Although his explanations was not correct (the tracks were in reality positive muons from pion decays at rest). The experimental discovery of juCF was achieved at the end of 1956 in Berekely by L.W.Alvarez team looking at bubble chamber pictures [2J. A muon (ju) is a leptonic elementary particle and has a finite lifetime of 2.2jus .Since the mass of a muon 207 times larger than an electron, the size of an exotic atom/molecule containing the negative muon is much smaller than an electronic atom/molecule .When the negative muon binds to hydrogenic nuclei (poroton,p,deuteron,d,or triton,t) likcff^ ,a nuclear fusion reactions occurs in the muonic molecular ion, for example {dt/j)+^He + n + /u~ (1) The muon does not take part in the nuclear reaction directly but only catalyzes the reaction. This process the muon catalyzed fusion. The pt reaction is one pf the least known of all processes of /JCF in the mixture of hydrogen isotopes. It is very important to gain information on reaction characteristics of all muonic processes in HT mixture(e.g., the rate of muon transfer from pju atom to triton ,the rate of transition between hyperfine levels of tfi atoms ,the rate of formation of the ptju molecule ,and the rate of nuclear synthesis in it) to interpret correctly the results of experiments in the triple mixture of hydrogen isotopes H-D-T and to describe the kinetics of all processes occurring in the mixture. From the theoretical point of view, the experiments investigating juCF processes in hydrogen -tritium mixture will allow one to test an algorithm describing a three-body system of particles interacting according to coulomb rule. It is necessary to emphasize the importance of the /JCF study in HT mixture in order to obtain the information about characteristics of pt -reaction at ultra law energy range (« KeV).The investigation of the reaction between light nuclei at ultra law energies (« KeV) is very important for verification of fundamental symmetries in string interactions [3-5],the contribution of muon exchange currents [6-9] and to solve some astrophysical problems [l 0-12] .At present ,there are few experiments [13 —17] that investigate characteristics of juCF in a H-T mixture. Only one [l3] was performed with a HT mixture and the references [14-17] with triple mixture H-D-T. In this paper the authors survey the major areas of research: Section I describes the details of the kinetics of the juCF in HT mixture, while Sec.II describes the nonlinear point dynamics equations in HT mixture .Sec.III discusses on the numerical solution of these equations versus time in muon life time range. Sec.IV describes the optimum cycling coefficient and energy gain. Our calculations show that ,the optimum value of muon cycling coefficient at Ct - 0.01 (tritium concentration) is equal to 106 .The experiments with the HT mixture at the meson facility PSI (Switzerland) are

237 13 International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye planned to be optimized to gain the precise information about the desired juCF .In this paper our obtained results from theoretical calculations and experimental results are compared with togetherand we can receive that the obtained results are in good agreement with measured values.

Key words: muon cycling, HT mixture, mu-atomic, mu-molecular, fusion.

References: 1. F.C.Frank, Nature 160,525(1947). 2. L.W.Alvarez, H.Branda, F.S.Craford et al, Phys.Rev.105, 1125 (1957). 3.S.P.Merkuriev,et al.,Proc.Int.Con.on the Theory and Few Body and Quark-Hadronic Systems ,Dubna,D4-87-692,6( 1987). 4. J.L.Friar, Proc.Int. Con.on the Theory and Few Body and Quark-Hadronic Systems, Dubna, D4-87-692, 70(1987). 5. H.Paetzgen, Schieck,Few Body Systems,5,171(1988). 6. C.Bargholz, Nucl.Phys.A,474,l(1987). 7. J.L.Fariar,Phys.Lett.B,251,l 1(1990). 8. J.Torre, B.Goulard, Phy.Rev.C.28, 529(1983). 9. A.CPhillips, Nucl.Phys.A 184,337(1972). 10. C.Rolfs, Proc.Int.School of Phys."Enrico Fermi", Course C.3, Monastero, 23 June -3 July 1387.p.417. 11. J.N.bachcall, M.H.Pinsonneault, Rev.Mod.Phys.64, 885(1992). 12.M.Arnould,M.Forestini,Nuclear Astrophysics,Proc.of the Rhird Int.Summer School ,La Rabida, Huelve,Spain,Research Reportsin Physics(Springer-Verlag,1988),p.48. 13. F.J.Hartmann et al.,Muon Catal.Fusion 2,53(1988) . 14. P.Baumann et al., Muon Catal.Fusion 5, 87(1990); F.J.Hartmann et al., Hyp.Int.82,259(1993)., 15-F.Mulhauser et al.Pysc.Rev A 53, 3069(1996). 16- F.Mulhauser et al, Hyp.Int.l 19,35(1999). 17-V.R.Bom et al., Hypfine Interact.118, 103(1999).

238 TR0700405

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

FRACTAL REACTOR: AN ALTERNATIVE NUCLEAR FUSION SYSTEM BASED ON NATURE'S GEOMETRY Todd Lael Siler Psi-Phi Communications, LLC 4950 S. Yosemite Street, F2-325 Greenwood Village, Colorado 80111 USA E-mai 1: [email protected]

ABSTRACTS The author presents his concept of the Fractal Reactor, which explores the possibility of building a plasma fusion power reactor based on the real geometry of nature [fractals], rather than the virtual geometry that Euclid postulated around 330 BC(1); nearly every architect of our plasma fusion devices has been influenced by his three-dimensional geometry. The idealized points, lines, planes, and spheres of this classical geometry continue to be used to represent the natural world and to describe the properties of all geometrical objects, even though they neither accurately nor fully convey nature's structures and processes. (2) The Fractal Reactor concept contrasts the current containment mechanisms of both magnetic and inertial containment systems for confining and heating plasmas. All of these systems are based on Euclidean geometry and use geometrical designs that, ultimately, are inconsistent with the Non-Euclidean geometry and irregular, fractal forms of nature (j). The author explores his premise that a controlled, thermonuclear fusion energy system might be more effective if it more closely embodies the physics of a star. This exploratory concept delves into Siler's hypothesis that nature's star "fractal reactors" are composed of fractal forms and dimensions that are statistically self-similar, (4) as shown in Figures 1 & 2. In effect, their form determines their function.

Exact Self-Similarity Statistical Self-Similarity Surface of the Sun (Fractal dimensions) (Fractal dimensions) (Fractal dimensions)

Figure 1, Comparison of different kinds of natural fractal images. Photos from Richard P. Taylor article, "Order in Pollock's Chaos," in Scientific American, December 2002; p.l 18.); "A Brief History of Chaos." DOT photo of the Sun's active region AR10786; the field measures 182 x 133 arcsec. (The inserted photo of Earth shows the scale).

239 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Classical Geometry Euclidean Dimensions and Reduction

0

Simple Fractal Geometry Fractal Scales and Reduction

Complex and Simple Complex and Simple

Figure 2. Chart comparing fractal and classical geometries. (Inspired by and based on Rhonda Roland Shearer's article "From Flatland To Fractaland: New Geometries In Relationship To Artistic And Scientific Revolutions," in Fractal Geometry and Analysis, The Mandelbrot Festschrift, C.J.G. Everts/-, H. 0. Peitgen and R. F. Voss, eds., [World Scientific Publishers, July 1996]; 617-625; also "Real Or İdeal? DNA Iconography In A New Fractal Era," Art Journal 55, no. 1, College Art Association [Spring 1996]: 67)

The Fractal Reactor concept uniquely combines both magnetic and quasi-inertial confinement mechanisms (see Figures 3-5). The fractal magnets are designed to approximate the gravitational forces in a star that contribute to the compressional heating of plasmas in conjunction with the common dynamo phenomenon; this event commonly occurs when the self-organizing, superheated plasmas generate their own intense magnetic fields, which, in turn, sustain their high-temperatures and density. Siler's concept aims at reverse engineering and re-creating the power of the Sun on a relativistic scale, first considering the possibilities of applying fractal geometry to help improve the effectiveness of the containment systems and vessels of fusion machines. This alternative approach to controlling the forces that govern plasmas aims at working with nature and not against it. Instead of using excessive brute force to jam a square peg (i.e., Euclidean geometry) into a round hole (i.e., fractal geometry), the ideas adventured here envision a way that fusion engineers could exert intense forces on the plasmas that approximate the gravitational forces in a star and that initiate the essential dynamo phenomenon.'-3'1 Grasping the physics of this basic phenomenon, which has been observed in the (6) and Spherical Torus(7), may provide key insights into heating plasmas using a combination of confinement mechanisms.(8) In examining the dynamo phenomenon, Dr. David Hill, one of the leaders of the Sustained Spheromak Physics Experiment (SSPX) at Lawrence Livermore commented: "...the

240 13n International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

main magnetic fields are generated by the plasma itself. It's a physical state the plasma wants to make naturally.. .The necessary strong magnetic fields are generated inside the plasma [by the magnetic dynamo]. In this regime, the plasma-fast-moving, super hot ions and electrons-produces its own confining magnetic fields. The magnetic fields pass through the flowing plasma and generate more plasma current, which in turn reinforces the magnetic fields...The dynamo drives the configuration [of fields and currents] towards a stable, minimum-energy state."(9)

Forts for Neutral Beam Injection

Sphı

Biological Shield (protection from plasma leaks)

Figure 3. A cut-away view of the roughly spherical-shaped Fractal Reactor. Note that the spicule-like ports for the neutral beam injection may be significantly fewer in number and more loosely organized than shown here. Also, these elements might be incorporated in the Ohmic heating primary windings and electromagnets. This drawing shows the general shapes of the integrated, magnetic and quasi-inertial confinement systems for this alternative, Controlled nuclear fusion device. (Drawing by Todd Siler.)

241 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Plasmas

Biological

Bursts of charged Flexible, moveable, fractal magnets and quasi- \ particles issuing from inertia! confinement mechanisms (wirele« l Npiltral Rpflm *ni™tnr Lithium blanket, or "gas blanket" (inside the Fractal

Figure 4. 3D visualization of the Fractal Reactor concept. This cut-away view shows the interconnected, irregular- shaped fractal-like, super-conducting magnets that continually move and oscillate, creating roughly spherical wave- like [minima/maxima] magnetic fields. The plasma is wide ranging in terms of its density, temperature, time/duration and confinement. (Model fabricated by Roger Leitner, based on Siler's sculptural sketches.) Note: the superconducting magnets move at different rates, generating variances in the movement of magnetic currents of plasma, and initiating the dynamo phenomenon. One alternative design for the fractal magnets is the Nobel-Prize winning "BuckyBall," or Icosahedral Fullerene C54O - only fracta/ized. The magnets could be carefully controlled, increasing and decreasing the pressures and temperatures of plasma within the core of the Fractal Reactor - thus controlling the degrees of compacting the hydrogen isotopes deuterium and tritium. Note that this device would not be limited to the D-T fuel reactions exclusively.

242 13lh International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

Neutral Beam Injection Lithium Plasma Fractal Figures Sa&b. Top: image attempts to show the motions of the superconducting magnets in the Fractal Reactor and the roughly spherical fractal magnetic fields they generate. The 3D computer-visualization was created the Anark Corporation, a media production company. Bottom: This interpretive drawing shows the interactions of these magnets with the piasmas they generate. It also presents a possible Prototype of the Fractal Reactor magnetic confinement system. This image of a Sunflower structure could serve as the general design for the fractal superconducting magnets Note the self-similarity of the elements. (Drawing inspired by and based on Jay Kappraff, Connections: The Geometric Bridge Between Art and Science.

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243 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, İstanbul, Türkiye

The Fractal Reactor concept uses the physics of the Sun as its principal model and prime source of design principles. It creatively explores ways we can learn about how our current magnetic confinement and inertial fusion devices are analogous to the systems operating within a star. In doing so, this creative inquiry challenges the fusion industry to consider more realistic and naturalistic ways of representing the designs and dynamics of controlled fusion energy systems. Siler wonders whether this fundamental issue concerning our classical geometrical designs for the containment vessels and confinement mechanisms may be hindering rather than helping our success in sustaining and harvesting hot plasmas for energy purposes. Perhaps, by experimenting with fractals in nuclear fusion technology, we may be better equipped to generate and manage high performance plasmas. He uses the Fractal Reactor concept as a catalyst for discussing the possibility of redesigning the vessel geometry of a plasma fusion device, in an effort to make it more effective and efficient system. This presentation will be accompanied by an interactive, 3D visualization that broadly shows how the principles of statistical self-similarity and scaling invariance, which are central to fractal geometry, may be applied to plasma physics. This computer-animation also helps differentiate the Fractal Reactor concept from the Spheromak, Spherical Torus, Field-Reversed Pinch, along with various Inertial Laser Fusion devices that relate to aspects of this concept and approach. The visual suppositions and premises posed by the Fractal Reactor concept aim to spark innovative thinking on how we can consistently generate, control, and sustain thermonuclear teactions in a way that more accurately replicates nature's way—including the way it manages unstable, self-organizing plasmas with radical temperature gradients. To this end, the author advocates applying common sense, observational science, and reverses engineering, <-I0) in re-creating the dynamics of a star. In attempting to move beyond his symbolic Artist Concepts, physical analogies and visual suppositions, Siler has begun the process of composing a compv.cational model. When completed, this mathematical model will demonstrate how the irregular-shaped, fractal, superconducting magnets and quasi-inertial confinement mechanisms in his new system will work in real-life. The author provides an overview of his work to date, and his plans for advancing the development of this work by collaborating with plasma physicists and fusion specialists who can create the mathematical models. Using various mathematical tools, including nonlinear partial differential equations, the models will demonstrate and describe:

(1) The actual movement and behavior of the roughly spherical, fractal superconducting magnets and the complex magnetic fields they generate; (2) The nuclear physics of the self-organizing plasmas that are generated and confined in this new fusion system; and (3) The motion of the fractal magnets in the Fractal Reactor and their interactions with the high-temperature plasmas they generate.

These computational models step us towards creating a proof of principle Fractal Reactor device, revealing the dynamics of this new system. Furthermore, they can help the fusion community gain a deeper understanding of the behavior of self-organizing plasmas, such as those observed in the magnetic dynamo phenomenon. To this end, the models can contribute to constructing—and eventually conducting—the empirical studies that can lead to some breakthroughs with tangible, practical applications for the nuclear fusion energy industry. The author has been advised by a number of outstanding plasma physicists who sense that it may be necessary to "create some new kinds of mathematics that can marry hydrodynamic and Maxwell's equations in accommodating

244 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye the Fractal geometry and fractal boundary conditions" that are core to the Fractal Reactor concept/11-1 The purpose of this presentation at ICENES is to respectfully question the path plasma fusion community has taken over the past fifty years, and to try to engage the adventurous mathematical physicists and fusion specialists attending the conference to be open to this concept and its implications. More importantly, I'm hoping that the concept is intriguing enough to attract the attention of innovators who might consider collaborating on composing a mathematical model that meets the bold challenges posed by realizing this technological innovation. In reviewing the Fractal Reactor concept, Fred Alan Wolf, (Ph.D., Theoretical Physics and Plasma Physics from UCLA, 1963), author of the American Book Award for Science Taking the Quantum Leap (1982), has written: "In the design of today's fusion devices it's the function that determines the form. By that I mean that plasma physicists know how to solve the usual Euclidian geometric (in my day Vlasov-Maxwell) equations of plasma physics and magneto hydrodynamics and so tend to design machines that their equations can describe. Hence they can solve the theoretical problem, because their minds have been formed to follow the functions of classical science and math. Unfortunately, nature doesn't work that way. It seems we have turned the problem upside down—it should be form determining function as the Fractal Reactor concept indicates. In other words we can see what Nature designs so why not follow her?" (-12)

REFERENCES AND NOTES

1. Review Euclid's classic book Elements, which was written at the time of Ptolemy I. 2. Note that in Phaenomena, Euclid relates his spherical geometry to render celestial objects studied in astronomy. T. Phillips, "The Transparent Sun" (Science@NASA); A group of applied mathematicians at the St. Andrews Solar Magneto Hydrodynamics (MHD) Theory Group, Scotland, is studying the nonlinear interaction between the Sun's magnetic field and its plasma interior or atmosphere. Note: their mathematical modeling techniques are informed by observational data from satellites, such as SoHO, Yohkoh and TRACE, as well as ground- based observatories, such as Kitt Peak. 3. T. Siler, "Fractal Reactor: A New Geometry for Plasma Fusion," in Proceedings for the 3rd Symposium on Current Trends in International Fusion Research: Review and Assessment (1999). A. Bunde and S. Havlin, (eds.), Fractals and Disordered Systems. 2nd rev. (Springer, 1996); K. Falconer, Fractal Geometry. (Wiley, 1990), and Fractals: Non-Integral Dimensions and Applications. (Wiley, 1991); IFIP Conference on Fractals in the Fundamental and Applied Sciences (North-Holland /Elsevier Science Publishers, Co., 1991); L. Nottale, Fractal Space-Time and Microphysics: Towards a Theory of Scale Relativity. (World Scientific, 1993); R.P. Taylor, "Order in Pollock's Chaos," in Scientific American, December 2002; p. 118. 4. J. Briggs, Fractals: The Patterns of Chaos. (Simon & Schuster, 1992); B. Mandelbrot, The Fractal Geometry of Nature. (W.H. Freeman, 1982); J. Kappraff, Connections, The Geometric Bridge between Art and Science. (McGraw-Hill, 1991); I. Toth, "Non-Euclidean Geometry before Euclid," in Scientific American (November 1969), pp. 87-98. Also read J. Gleick, Chaos; Making a New Science. (Penguin Books, 1987); M.R. Schroeder, Fractals, Chaos, Power Laws: Minutes From An Infinite Paradise. (W.H. Freeman, 1991).

245 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

5. Forest et al., Madison Dynamo Experiment (MDX), University of Wisconsin-Madison, 1999; J. Sherwood, "Magnetic Dynamo Appears to Shape Planetary Nebulae," UniSci. Daily University Science News (Jan. 25, 2001). 6. K.I. Thomassen et al, "The Spheromak Path to Fusion Power," in Proceedings for the 3rd Symposium on Current Trends in International Fusion Research: Review and Assessment, Washington, D.C., 8-12, March 1999. 7. M. Peng, "Cost-Effective Spherical Torus Steps Toward Fusion Power," in Proceedings for the 3rd Symposium on Current Trends in International Fusion Research: Review and Assessment, Washington, D.C., 8-12, March 1999. 8. T.A. Heppenheimer, The Man-Made Sun: The Quest for Fusion Power. (Little Brown and Company, 1982) 9. A. Heller, "Experiment Mimics Nature's Way With Plasmas"; D. Hill et. al work on Lawrence Livermore's Sustained Spheromak Physics Experiment (SSPX), in collaboration with the Sandia and Los Alamos national laboratories. Also read J. Herrera, "The Self-Organization Concept in Magnetic Confinement Fusion," in Proceedings for the 3rd Symposium on Current Trends in International Fusion Research: Review and Assessment, Washington, D.C., 8-12, March 1999. 10. The Society of Manufacturing Engineers defines Reverse Engineering as, "the process of taking a finished product and reconstructing design data in a format from which new parts or molds can be produced." In this instance, the finished product is the Sun; its design data has been gleaned by many decades of empirical studies and mathematical models informed by astrophysics. Also, the Military Handbook MIL-HDBK-115 defines Reverse Engineering as "...duplicating an item functionally and dimensionally by physically examining and measuring existing parts to develop the technical data (physical and material characteristics) required for competitive procurement." 11. This suggestion was offered by J.C. Sprott, Department of Engineering Physics at University of Wisconsin—Madison. Professor Jay Kappraff in the Mathematics Department at New Jersey Institute of Technology, emailed me: "Your fractal reactor sounds very interesting. I recommend that you look at the work of N. Rivier on froths since they form natural fractal like patterns like the ones you illustrate. The idea of connecting plasmas to natural system has the right ring to it. However, it will be quite a task to wed the hydrodynamic and Maxwell's equations to your geometry. It seems as though some new kinds of mathematics may have to be used to accommodate fractal boundary conditions." In 1974, Dr. Kappraff did his doctoral work at New York University's Courant Institute in magneto hydrodynamics applied to controlled fusion. He was very active with the fusion group led by Dr. Harold Grad, and performed the first bifurcation of a plasma equilibrium applied to the Stellarator. Dr. Kappraff s thoughts on how to proceed with developing the Fractal reactor concept were further echoed by another excellent plasma physicist, Dr. Manos Chaniotakis at MIT's Fusion Lab, who was very insightful, too. Dr. Chaniotakis suggested a couple of viable ways to advance my project with the creation of the three separate but interrelated computational models I've listed. 12. This quote is from an email correspondence with the plasma physicist and author, Fred Alan Wolf, Ph.D.; it is reprinted here courtesy of Dr. Fred Alan Wolf.

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THERMONUCLEAR FUSION BY MECHANICAL ADIABATIC COMPRESSION OF A DENSE PLASMA

David W. Kraft Division of Natural Sciences and Mathematics, University of Bridgeport Bridgeport, CT 06601, USA E-mail: [email protected]

ABSTRACT Thermonuclear fusion rates for particles of a single species are proportional to n2 where n are the number density of the reacting particles. Standard magnetic confinement techniques employ relatively thin plasmas and therefore require temperatures of the order of 108 K. We propose a method to exploit the n2 factor and hence to attain appreciable fusion rates at much lower temperatures. Principal features include the rapid adiabatic compression by a piston of a dense gas of deuterium in an adiabatically insulated reaction chamber and a reduction in the degrees of freedom of the plasma particles such as may be effected by an electric discharge or by application of magnetic fields. Model calculations consider the adiabatic compression of one mole of deuterium initially at room temperature and pressure and the resulting fusion energy release under conditions of varying degrees of freedom is compared with the work done to compress the piston. Additional fusion enhancement factors and the timing of the reduction of the degrees of freedom within the compression interval are discussed. The calculations contain simplifying assumptions and idealizations and hence the results may be regarded as limiting targets to be attained under ideal conditions.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

BENCHMARKING THE CAD-BASED ATTILA DISCRETE ORDINATES CODE WITH EXPERIMENTAL DATA OF FUSION EXPERIMENTS AND TO THE RESULTS OF MCNP CODE IN SIMULATING ITER

Mahmoud Z. Youssef University of California-Los Angeles, Los Angeles, CA 90025, E-mail: [email protected]

ABSTRACT Attila is a newly developed finite element code based on Sn neutron, gamma, and charged particle transport in 3-D geometry in which unstructured tetrahedral meshes are generated to describe complex geometry that is based on CAD input (Solid Works, Pro/Engineer, etc). In the present work we benchmark its calculation accuracy by comparing its prediction to the measured data inside two experimental mock-ups bombarded with 14 MeV neutrons. The results are also compared to those based on MCNP calculations. The experimental mock-ups simulate parts of the International Thermonuclear Experimental Reactor (ITER) in-vessel components, namely: (1) the Tungsten mockup configuration (54.3 cm x 46.8 cm x 45 cm), and (2) the ITER shielding blanket followed by the SCM region (simulated by alternating layers of SS316 and copper). In the latter configuration, a high aspect ratio rectangular streaming channel was introduced (to simulate steaming paths between ITER blanket modules) which ends with a rectangular cavity. The experiments on these two fusion-oriented integral experiments were performed at the Fusion Neutron Generator (FNG) facility, Frascati, Italy. In addition, the'nuclear performance of the ITER MCNP "Benchmark" CAD model has been performed with Attila to compare its results to those obtained with CAD-based MCNP approach developed by several ITER participants. The objective of this paper is to compare results based on two distinctive 3-D calculation tools using the same nuclear data, FENDL2.1, and the same response functions of several reaction rates measured in ITER mock-ups and to enhance confidence from the international neutronics community in the Attila code and how it can precisely quantify the nuclear field in large and complex systems, such as ITER. Attila has the advantage of providing a full flux mapping visualization everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. In addition, the turnaround time for an Attila run is relatively shorter than MCNP and as such, Attila lends itself to be a powerful calculation tool for fusion components design in which frequent changes are made.

Favorable results are obtained with Attila in the experimental benchmarking exercise. For example, in the W-experiment, the results show that Attila can predict the very high threshold reaction [Zr90 (n, 2n), Ni58 (n, 2n)] within -5-8% as compared to -15-20% with MCNP. There is under prediction for Nb93 (n, 2n), Al27 (n, a) and Fe56 (n, p) reactions by -5-15% (MCNP also under predict these reactions by -5-25%). However, for low energy reactions such as Au197(n,g), Attila shows larger underestimation by -25% than MCNP results which give calculated-to experimental (C/E) value of- 1-1.09. These comparisons and the result from benchmarking Attila using the ITER MCNP CAD model will be presented in this paper.

248 PARALLEL SESSION 14B: NUCLEAR TECHNIQUES . — —. Battı mut |||U | TR0700408

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, İstanbul, Türkiye

SUPERCRITICAL CARBON DIOXIDE BRAYTON POWER CONVERSION CYCLE FOR BATTERY OPTIMIZED REACTOR INTEGRAL SYSTEM

Tae W. Kim, Nam H. Kim, Kune Y. Suh* Seoul National University, San 56-1 Sillim-dong, Gwanak-gu, Seoul, 151-744, Korea E-mail.-kysuh@snu. ac. kr

ABSTRACT Supercritical carbon dioxide (SCO2) promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. The SCO2 Brayton cycle efficiency as high as 45% furnishes small sized nuclear reactors with economical benefits on the plant construction and maintenance. A 23 MWn, lead-cooled Battery Optimized Reactor Integral System (BORIS) is being developed as an ultra-long-life, versatile-purpose, fast-spectrum reactor. BORIS is coupled to the SCO2 Brayton cycle needing less room relative to the Rankine steam cycle because of its smaller components. The SCO2 Brayton cycle of BORIS consists of a 16 MW turbine, a 32 MW high temperature recuperator, a 14 MW low temperature recuperator, an 11 MW precooler and 2 and 2.8 MW compressors. Entering six heat exchangers between primary and secondary system at 19.9 MPa and 663 K, the SCO2 leaves the heat exchangers at 19.9 MPa and 823 K. The promising secondary system efficiency of 45% was calculated by a theoretical method in which the main parameters include pressure, temperature, heater power, the turbine's, recuperators' and compressors' efficiencies, and the flow split ratio of SCO2 going out from the low temperature recuperator. Development of

Modular Optimized Brayton Integral System (MOBIS) is being devised as the SCO2 Brayton cycle energy conversion cycle for BORIS. MOBIS consists of Loop Operating Brayton Optimization Study (LOBOS) for experimental Brayton cycle loop and Gas Advanced Turbine Operation Study (GATOS) for the SCO2 turbine. Liquid-metal Energy Exchanger Integral System (LEXIS) serves to couple BORIS and MOBIS. LEXIS comprises Physical Aspect Thermal Operation System (PATOS) for SCO2 thermal hydraulic characteristics, Shell-and-tube Overall Layout Optimization Study (SOLOS) for shell-and-tube heat exchanger, Printed-circuit Overall Layout Optimization Study (POLOS) for printed circuited heat exchanger, and Flat-plate Integrated Layout Optimization Study (FILOS) for plate fin heat exchanger. PATOS has been conceptually designed to set up the experimental loop. The operating conditions include pressure 7.4-8.0 MPa, bulk temperature 24~40°C, wall temperature 80-160°C, heat flux 20-80 kW/m2, and mass flow rate 0.01-0.1 m/s in tube diameter 1.27 cm. GATOS is developed to achieve a high efficiency of 90% resorting to computational analysis. The hub radius is 40 cm, and the tip radius is 46.5 cm. The blade height is 6.5 cm, and the mean radius is 43.5 cm. Both angles of the camber inlet and outlet are 60°. The chord and leading edge are 8 and 0.1 cm long. The boundary conditions are the inlet total pressure 20 MPa at temperature of 823 K, revolution of rotor 60 per second, and the average static pressure at the outlet 17 MPa.

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UTILIZING THE SLOWING-DOWN-TIME TECHNIQUE FOR BENCHMARKING NEUTRON THERMALIZATION IN GRAPHITE

T. Zhou, A. I. Hawaii*, B. W. Wehring Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695, USA E-mail [email protected]

ABSTRACT Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in "reacor grade" graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70cm x 70cm x 70cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multi- channel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured values. At room temperature, the results indicate a deviation of nearly 20%-30% in the thermal energy range. This deviation is attributed to the inadequate representation of graphite in the current data libraries.

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SENSITIVITY ANALYSIS OF CORROSION PRODUCT ACTIVITY IN PRIMARY COOLANT OF PWR

Muhammad Rafique3'*, Nasir M. Mirzab, Sikander M. Mirzab ^Department of Physics AJK University, Muzaffarabad, 13100, Azad Kashmir, Pakistan b Department of Physics & Applied Mathematics, Pakistan Institute of Engineering & Applied Sciences, P. O. Nilore, Islamabad 45650, Pakistan E-mail: [email protected]

ABSTRACT The objective of this paper is to explicit the behavior of different radiochemical species in the primary coolant circuit of pressurized water reactor (PWR) plants. For this purpose a detailed study based on the sensitivity analysis of activated corrosion products in primary coolant of a typical PWR have been carried out under normal reactor operations. Main purpose of this study is to identify the sensitive parameters, (i.e. that cannot be changed without changing the optimal solution). Different parameters like, lk (rate at which primary coolant water loses its water from k'th leak), is: ( rate at which radioactive material is removed from scale on piping surfaces), Kc ( rate at which radioactive material is removed from scale on core surfaces ),scQc (removal rate from core surfaces) and £pQp (removal rate from piping surfaces), SjQs (ion-exchanger removal rates) appearing in the mathematical models have been monitored closely as the study is implemented. Computer program CPAIR-RC has been modified to accommodate for time dependent corrosion rates. Parametric study for three radioisotopes 56Mn, 58Co, and 60Co have been done. In the course of normal operation of reactor major part of corrosion product activity (CPA) comes from isotope 56Mn while 58Co and 60Co dominate the activity after shutdown of reactor. Parametric study suggests that the effect of ion exchanger removal rates on CPA is foremost. For removal rate of 300 cm3 per second the specific activity due to 56Mn has the maximum value of 0.96GCi/cm3 after 1000 hours of reactor operation and this value decreases drastically to 0.225GCi/cm3 at removal rate of 900 cm3 per second. In the same way sensitive 56 58 60 analysis for Kc suggests that CPA due to Mn, Co and Co varies significantly with the 56 58 60 variation in Kc values. At Kc=25 cmVsec the specific activity due to Mn, Co, and Co in primary coolant after 1000 hours of reactor operation have the values 0.327, 0.154, 3 3 0.0116DCi/cm respectively. Which decreases at higher values of Kc and at Kc=55 cm /sec the 3 new values are 0.315, 0.072, 0.0053DCi/cm . Variations in the values ofKp,epQp have trifling effect on CPA suggesting that these parameters are less sensitive as compared to others. Some of the results of CPAIR-RC have been validated against the code CRUDSIM/MIT. There is a fair agreement between activity ratios of 58Co to 60Co as predicted by CPAIR-RC and CRUDSIM/MIT codes for the optimal values chosen for the sensitive parameters appearing in the mathematical models. Author Keywords: sensitive analysis; optimal values, mathematical models, corrosion products; Pressurized Water Reactors (PWRs); sensitive parameters.

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ANALYSIS AND IMPROVEMENT OF CYCLOTRON THALLIUM TARGET ROOM SHIELD

jo N., 2Raisali G. 'Nuclear Science and Technology Research Institute (NSTRI),. Agriculture, Medicine and Industry Research School., SSDL and Health Physics Dept, Moazzen Blvd, Rajaee shahr, Karaj, Iran, P.O.Box: 31485-498 Tel: +98 261 4424073 Fax: +98 261 4464058 E-mail: [email protected] & [email protected]

2Gamma Irradiation Center, Atomic Energy Organization of Iran, End of Kargare Shomali, Tehran, Iran, P. O. Box: 11365-3486 Tel: +9821 88004065 Fax: +9821 88009054 E-mail: [email protected] & [email protected]

ABSTRACT Because of high neutron and gamma ray intensities during thallium-203 target bombardment, thallium target room shield and its improvement have been investigated. Leakage neutron and gamma-ray dose rates in various points behind the shield are calculated by simulating the transport of neutrons and photons using Monte Carlo MCNP4C computer code. By considering target room geometry, its associated shield, neutron and gamma rays source strengths and spectra, three designs for enhancing shield performance have been analyzed; A door as a shield in maze entrance, covering maze walls with layers of some effective materials and adding a shadow shield in target room in front of the radiation source, have been considered as the parallel to the maze. Dose calculations carried out for each kind of suggested shields separately for different materials and dimensions, then the shield with better than has been constructed and It has been found that the deviation between calculated and measured dose values after upgrading is less than 20%. Key words: Streaming; Cyclotron; Neutrons; Gamma rays; MCNP code.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

ENHANCEMENT OF NUCLEAR HEAT TRANSFER IN A TYPICAL PRESSURIZED WATER REACTOR BY NEW SPACER GRIDS

Mohammad Nazifi*, Mohammadreza Nematollahi Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz, Iran Phone/Fax: +98-711-6287500, E-mail: [email protected]

ABSTRACT The fuel element geometry typically used in nuclear reactor is rod bundle whose rod-to-rod clearance is maintained by grid spacer. The heat generated in the rod by nuclear reaction is removed by coolant, usually in turbulent flow. The coolant moves axially through the sub- channels. Fuel spacer grid affects the coolant flow distribution in a fuel rod bundle, and so spacer geometry has a strong influence on a bundle's thermal-hydraulic characteristics such as critical heat flux and pressure drop. An understanding of the detailed structure of the turbulent flow and heat transfer in the rod bundle, used especially as nuclear fuel elements, is of major interest to the nuclear power industry for their safe and reliable operation. The flow mixing devices on grid spacer would enhance the mixing rate between sub-channels and promote the turbulence in sub- channel. The present study evaluates the effects of mixing vane shape on flow structure and heat transfer downstream of mixing vane in a sub-channel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. Six new shapes mixing vane designed by the authors, are simulated numerically to evaluate the performance in enhancing the heat transfer, in comparison with commercialized split vane. Standard K-epsilon model are used as a turbulence closure model and periodic and symmetry condition are set as boundary conditions. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. It is conformed that the turbulence in the sub-channel was significantly promoted by spacer and mixing devices but rapidly decreased to a fully developed level approximately 10 time of hydraulic diameter downstream of the top of spacer. Ring type mixer showed a high enhancement in nuclear heat transfer among the other mixing devices. Also the results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value.

Keywords: Pressurized water reactor, mixing vane, spacer grid, turbulent heat transfer, computational fluid dynamic, standard k-epsilon model.

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CALIBRATION EXPERIMENTS OF NEUTRON SOURCE IDENTIFICATION AND DETECTION IN SOIL

N. V. Gorin, E. N. Lipilina, G. V. Rukavishnikov*, D. V. Shmakov, A. I. Ulyanov Russian Federal Nuclear Center VNIITF, Russia E-mail: Grigory Rukavishnikov, [email protected], [email protected] E-mail: Victor S Mamontov

ABSTRACT In the course of detection of fissile materials in soil, series of calibration experiments were carried out on in laboratory conditions on an experimental installation, presenting a mock-up of an endless soil with various heterogeneous bodies in it, fissile material, measuring boreholes. A design of detecting device, methods of neutrons detection are described. Conditions of neutron background measuring are given. Soil density, humidity, chemical composition of soil was measured. Sensitivity of methods of fissile materials detection and identification in soil was estimated in the calibration experiments. Minimal detectable activity and the distance at which it can be detected were defined. Characteristics of in a borehole mock-up were measured; dependences of method sensitivities from water content in soil, source-detector distance and presence of heterogeneous bodies were examined. Possibility of direction detection to a fissile material as neutron source from a borehole using a collimator is shown. Identification of fissile material was carried out by measuring the gamma-spectrum. Mathematical modeling was carried out using the PRİZMA code (Developed in RFNC-VNIITF) and MCNP code (Developed in LANL). Good correlation of calculational and experimental values was shown. The methodic were shown to be applicable in the field conditions.

256 TR0700414

13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

FUSION BY QED CONFINEMENT?

T. V. Prevenslik Foehrenweg 20, Berlin 14195, Germany Email: [email protected]

ABSTRACT Inertial confinement fusion (ICF) embodies the simultaneous firing of large-scale multiple high- intensity mega-joule laser pulses at targets of pea-sized deuterium pellets. Fusion is initiated on the premise that temperatures over 100 million degrees and density lOOOx normal solid density are produced in pellet implosion. Contrary to large scale ICF reactor designs, neutrons were recently found in Berlin with a significantly down-sized table-top system using a 815 nm laser depositing ~ 35 femto-second pulses of 120 milli-joules to a spray of submicron droplets of heavy water. Ionization of the D2O is thought to leave the submicron droplets with charged D2O+ ions as the electrons rapidly escape at which time the droplets undergo Coulomb explosions. Neutrons are produced in the collisions of the D2O+ ions with those emitted from neighboring droplets. Although a convenient source of submicron D2O droplets, the spray lacks containment of the Coulomb explosion to produce the high pressures necessary to significantly increase neutron yield. A containment shell encapsulating the D2O is proposed here to briefly increase the pressure by confining the D2O+ ions in the Coulomb explosion. In pea-sized targets this is not possible because the shell would vaporize under the high temperatures necessary for implosion. But with submicron targets, the temperature does not increase during the absorption of the laser photons because of the quantum electrodynamics (QED) confinement of the electromagnetic (EM) radiation.. Specifically, IR radiation from absorbed photons is precluded having half-wavelengths larger than the target diameter, or equivalently any increase in target temperature that accompanies the IR radiation is forbidden. What this means is that in pea-sized targets large in relation to the half-IR wavelength, the deposition of laser energy is allowed to be conserved by an increase in temperature; whereas, in submicron targets far smaller than the respective half-IR wavelength, any temperature increase is forbidden. Absent an increase in temperature, EM energy is conserved by increasing the frequency of the EM radiation within the target. By this theory, called cavity QED induced EM radiation, D-D fusion in a submicron target is initiated by the absorption of a prompt pulse of laser photons in a non-equilibrium process that increases the EM energy to the 10 keV level before thermal equilibrium produces the equivalent 108 K temperatures. In this way, fusion in submicron targets is initiated through the electronic states of the D2O molecule - not by the increasing the temperature as would be the case in pea- sized targets. In the D2O filled containment shell, the photons in the 815 nm laser pulse are absorbed by increasing their frequency within the D2O target to its EM resonant frequency, which for submicron D2O targets is to vacuum ultraviolet (VUV) levels. The D2O promptly ionizes with the VUV radiation produced in the target from the remainder of the laser pulse raising the ionized electronic states by multi-photon processes to levels of a few keV. Without an increase in temperature, the containment shell remains intact, at least long enough for the pressure from the D2O+ ion Coulomb explosion to initiate fusion at the keV electronic levels of the D2O molecules. Once fusion is initiated, the containment shell vaporizes as the high level of EM energy in the electronic states is conserved by high temperatures at thermal equilibrium. Alternative fuels to heavy water and various shell confinements are presented.

257 TR0700415

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

RELAP5 SCDAP RIA transients analysis

Dr. Gheorghe Negut, Institute for Nuclear Research, Pitesti, Romania Prof. Dr. Ilie Prisecaru Conf. Dr. Daniel Dupleac University Politehnica Bucharest, Bucharest, Romania

ABSTRACT Some Romanian research programs require the study of fuel behavior during transients- Romanian TRIGA Annular Core Pulse Reactor (ACPR) is provided with capabilities to produce short pulses that simulate Reactivity Initiated Accidents (RIA). A program was initiated to study CANDU type fuel in transient situations by using these reactor capabilities. Since 1984, over 54 tests were performed with fresh un-irradiated fuel specimens, placed in a special capsule. The goal of these tests was to determine the clad failure threshold and to determine fuel-clad-coolant interaction during transients.

The irradiation tests conditions for this special capsule are atmospheric pressure and environmental temperature. This capsule is placed in the ACPR central dry cavity during the tests.

During the pulse, in a very short time a significant energy is deposed in the fuel, which heats up the fuel. The clad temperature rises to over 1000 C. Due to the high temperature; the clad will quickly be covered with a vapor layer. Subsequent cooling of the clad is described as rewetting, involving thermal hydraulic phenomena which take place in the re-flooding phase following a LOCA accident, when on the fuel elements, covered with steam, cold water is injected from the emergency cooling system.

The capsule and the fuel specimen was modeled using RELAP5-SCDAP code. To document our SCDAP model we used some JAERI experiments. These evaluations show a good agreement.

The SCDAP model developed for clad temperature prediction for the tests in Romania's ACPR, with the atmospheric capsule and fresh fuel, as well as for similar JAERI reactor (Japan) tests proved to have good predictions and, also, gave a good prediction for clad failure threshold energy. The predicted failure threshold is confirmed both for the ACPR tests and for the JAERI tests.

258 TR0700416

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

FUSION CHANNEL OF pd CHARGE-SYMMETRIC ION INCLUDING PHOTONS

Gheisari, Rouhollah Physics Department, Persian Gulf University, Boushehr,Iran

ABSTRACT The charge- symmetric pseudo nucleus pd is formed in the cascade processes in the muon catalyzed fusion. The nuclear fusion in pdju ion can be considered in the photon field. For the spin states of pd (L=0) system, employing a new space wave function of three-body, the matrix element Ml proportional to

K He \M and the fusion rate J R|2F (2) for its ground state are calculated. The used wave function is introduced in the form of

(3)

The nuclear wave function % °'°(R)Yo,o is numerically calculated considering Wood-Saxon potential in the total Hamiltonian of the mentioned system. The good behavior of 9?(i?) is caused that our works are easily done in a short computation time. This function is linear from R =0 to 2.2 x IQ~10cm and then, is limited to 0.7068. The constant parameters of nuclear potential are obtained as well as those of the introduced wave function, when the boundary conditions are satisfied in our calculations. Notice that the notations {R,r) are Jacobean coordinates. The radiative pd fusion rates for the two spin states in the pdjx mesic molecule are found to be = 0A2jus~l and^ = 0.13/zs'1, close to experimental data. f, 2 /2

259

PARALLEL SESSION 15A: NUCLEAR FISSION POWER TR0700417

13lh International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

PRELIMINARY FEASIBILITY STUDY OF THE HEAT-PIPE ENHS REACTOR

Massimiliano Fratoni, Lance Kim, Sara Mattafirri, Robert Petroski, Ehud Greenspan

University of California at Berkeley, Berkeley CA 94720-1730 USA E-mail: [email protected]

ABSTRACT This preliminary study assesses the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor [1] to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE space nuclear reactor core [2], the HP-ENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The HPs extend beyond the core length and transfer heat to a secondary coolant that flows by natural- circulation. The HP-ENHS reactor is designed to preserve many features of the ENHS reactor including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walk-away passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor [1],

Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of possible advantageous features including: (1) significantly enhanced decay heat removal capability; (2) no positive void reactivity coefficients; (3) no direct contact between the fuel clad and coolant, hence, relatively lower wet corrosion of the clad; (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability.

The study focuses on four areas: material compatibility analysis, HP performance analysis, neutronic analysis and thermal-hydraulic analysis. Of four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the preferred working fluid and the HP working temperature is 1300 K.

The neutronic analysis found that it is possible to achieve criticality and maintain a nearly zero burn-up reactivity swing for at least 20 EFPY with a linear heat generation rate of 90W/cm. The preferred design uses nitride fuel made of natural nitrogen and loaded with depleted uranium and TRU from LWR spent fuel after a long cooling period and arranged in a very compact lattice of P/D=1.0.

The core is oriented horizontally with a square rather then cylindrical cross section for effective heat transfer to the secondary coolant. The preferred secondary coolant is LiF-BeF2. The HPs extend from the two axial reflectors in which the fission gas plena are embedded. For effective heat transfer to the secondary coolant it is sufficient to have the HPs extend less then 50 cm; the secondary coolant average outlet temperature is ~ I090K. The required reactor vessel height is significantly smaller than that of the reference ENHS: 9 vs. -20 m. The vessel diameter is slightly larger: 4 vs. ~3.5 m.

262 13U International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

References 1. E. Greenspan and the ENHS project team, "The Long-Life Core Encapsulated Nuclear Heat Source Generation IV Reactor," ICAPP'02, Holywood, FL, June 9-13, 2002. 2. D.I. Poston, "Nuclear design of the SAFE-400 space fission reactor", Nuclear News, December 2002.

263 TR0700418 i 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

OBSERVATIONS ON THE CANDLE BURN-UP IN VARIOUS GEOMETRIES

Walter Seifritz Mülacherstr. 44 CH-5212 Hausen/Switzerland E-mail: [email protected]

ABSTRACT We have looked at all geometrical conditions under which an autocatalytically propagating burn- up wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning spaghetti" reactor and the azimutally burning ,,pancake" reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometriy, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed.

A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology.

264 TR0700419

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

EXPERIMENTAL RESULTS FROM A REACTOR MONITORING EXPERIMENT WITH A CUBIC METER SCALE ANTINEUTRINO DETECTOR

Adam Bernstein Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94551, USA E-mail: Bernstein3@llnlgov

ABSTRACT Cubic meter scale antineutrino detectors can stably and no intrusively monitor both plutonium content and reactor power at the few percent level, at a standoff of a few tens of meters. Our Lawrence Livermore National Laboratory/Sandia National Laboratories collaboration has deployed a detector to demonstrate this capability at a 3 GWt pressurized water reactor in Southern California, operating 25 meters from the core center, and acquiring data over an approximate one year period. Such monitoring may be useful for tracking power output and plutonium buildup in nuclear reactors, constraining the fissile content and providing the earliest possible measurement of the amount of plutonium in the reactor core. We present our antineutrino event sample, and show that the observed change in antineutrino rate recorded in our detector over the reactor cycle correlates with plutonium ingrowth according to predictions. We present our current precision and estimate the attainable precision of the method, and discuss the benefits this technology may have for the International Atomic Energy Agency (IAEA) or other safeguards regimes.

600

Predicted rate Reported power • Observation, 24hr avg.

02/28/05 03/07/05 03/14/05 03/21/05 03/28/05 Date This figure shows the antineutrino data through a reactor turn-on, including a few day period of operation at 80% power. The left axis is the antineutrino rate; the right axis is the measured reactor power. A single overall normalization constant has been applied.

265 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

This figure shows LLNL/SNL detector, including the central liquid scintillator filled cells, the passive water shield and the muon.

266 TR0700420

13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye

FEASIBILITY STUDY OF SMALL LONG-LIFE WATER COOLED THORIUM REACTORS (WTRS) FOR PROVIDING SMALL QUANTITY OF ENERGY DEMANDS

Ismail1, Peng Hong LIEM, Sidik PERMANA, Hiroshi SEKIMOTO Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 Nl-17 O-okayama, Meguro-ku, Tokyo 152-8550, Japan. Tel./Fax. +81-3-5734-2955, E-mail: [email protected]

ABSTRACT A small scale, 30 ~ 300 MWth, of water cooled thorium reactor (WTR) is intensively investigated for supplying small quantity of energy demands in many remote and less developed areas of some developing countries. To provide suitable design for the areas, the reactors should have good design features. In the present study, both small-size and long core life features are chosen to be basic design goals of the reactors. Water cooled reactor type is used as a basic investigation design because it is a good proven nuclear power reactor technology. Thorium is introduced in this study to achieve some merits, such as higher conversion ratio and higher discharge fuel burnup compared to uranium fuel to provide basic characteristic for achieving good small long life reactors performances. The investigation covers some important parameters, i.e. enrichment, moderator to fuel ratio, discharge fuel burnup {physics aspect); and coolant void reactivity coefficient {safety aspect).

267

PARALLEL SESSION 15B: MISCELLANEOUS 13th International Conference on Emerging Nuclear Energy Systems June 03-08,'. TR0700491

METAPHYSICS METHODS DEVELOPMENT FOR HIGH TEMPERATURE GAS COOLED REACTOR ANALYSIS

Volkan Seker, Thomas J. Downar Purdue University Nuclear Engineering Building 400 Central Drive West Lafayette, IN,47907 E-mail: [email protected]; [email protected]

ABSTRACT Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology roadmap. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark require full three dimensional modeling and will be analyzed to include the malfunctioning of one of the de-fueling chutes and blockage of the helium flow channels in the side reflector at the PBMR-400 model.

270 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 200' TR0700422

INVESTIGATION OF TRITIUM AND 233U BREEDING IN A FISSION-FUSION

HYBRID REACTOR FUELLING WITH ThO2

Kadir Yıldıza, Necmettin Şahin8, H. Mehmet ŞAHİNb, Şenay Yalçın0, Taner Altınokd,Adem Acırb, Mustafa Bayrak6, Mahmut AIkane, Okhan Durukanf

^ Aksaray University, Engineering Faculty, Aksaray - TÜRKİYE (kyildiz@Niğde.edu.tr) b) Gazi University, Technology Faculty, Beşevler - Ankara - TÜRKİYE c) Bahçeşehir Üniversitesi, Mühendislik Fakültesi, Beşiktaş, İstanbul, Türkiye d)Kara Harb Okulu, Savunma Bilimleri Enstitüsü, ,Ankara 06654, Türkiye e) Niğde University, Engineering Faculty, Niğde - TÜRKİYE 0 Niğde Üniversitesi,, Natural Science Entitute, Niğde - TÜRKİYE

ABSTRACT In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 233U isotope which has very high quality fission cross-section with thermal neutrons. 233U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fusil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with TI1O2 for At=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons.(14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following,

2D + 3T -*4He (3.5 MeV) + n (14.1 MeV) (1)

The fuel zone made up of natural-ThO2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li2BeF4, LiF-NaF-BeF2, Li2oSngO, natural Lithium and \A\i?b^, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Lİ3N, Li2O, Li2O2, Li2Ti03, Lİ4SİO3, Li2ZrO3, LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S8-P3 approximation with Gaussian quadratures using the 238 groups library, derived from ENDF/B-V. For a self- sustained fusion fuel supply, a tritium breeding rate (TBR) > 1.05 is required. At the beginning of the reactor period (BOL) and at the end of the reactor period (EOL), LinPbs3 moderated blanket has well results compared to other moderators for TBR. According to

Lithium compounds used tritium breeding zone, at the BOL and at the EOL, LiH, L13N, Li2O, 233 Li2O2 and Lİ4SİO4 have well results. In the fuel zone of the fission-fusion hybrid reactor, U 232 233U isotopes are produced with Th (n,y) reactions. At the BOL and at the EOL, Lii7Pb83 moderated and Li2Zr03 tritium breeder blanket has well results compared to other moderators and tritium breeders for U breeding rate (UBR).

271 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, TR07nf\A91

NUMERICAL ANALYSIS OF STEADY STATE FLUID FLOW IN A TWO- DIMENSIONAL WAVY CHANNEL

Mofid Gorii1 and E. Hosseinzadeh2 Department of Mechanical Engineering, School of Engineering,1 University of Mazandaran, Babol, Iran E-mail: gorji@nit. ac. ir se_elham@yahoo. com

ABSTRACT Summary A simple geometry of the flow passage that may be used to enhance the heat transfer rate is called wavy and periodic channel. Wavy channel can provide significant heat transfer augmentation and was always important for heat transfer engineering and so far many researches have been done in this field.

In this paper, the effects of channel geometry and Reynolds number on the heat transfer coefficient, heat flux and pressure drop for the laminar fully developed flow in a two- dimensional channel whereas the walls are considered fix temperature is numerically investigated. The problem is solved for channel with one and two wavy walls and comparisons with the straight channel, in the same flow rate, have been performed. Results indicate that, by decreasing the channel wave length and the distance between the channel walls the pressure drop, heat flux and heat transfer coefficient increase.

Results and Conclusions

The following conclusion may be drawn:

1. In a specified channel, for the fluid flow with the constant Reynolds number, by decreasing the wave length from 0.2m to 0.1m, the pressure drop, heat flux and heat transfer coefficient increase by 37% , 54% and 29% respectively, whereas by decreasing the wave length from the same value the above mentioned parameters decrease to 108% , 143% and 47% respectively.

2. In a specified wave length, where the amplitude and the Reynolds number is constant, by increasing the distance between the walls from 0.15m to 0.25m, the pressure drop, heat flux and heat transfer coefficient decrease by 41% ,8% and 7.8% respectively.

References [1] J.C. Burns, T. Parks, J. Fluid Mesh, 29(1967), 405-416. [2] J.L. Goldestein, E.M. Sparrow, ASME J. Heat Transfer, 99 (1977), 187. [3] J.E.O. Brain, E.M. Sparrow, ASME J. Heat Transfer, 104 (1982), 410 [4] N. Sanie, S. Dini, ASME J. Heat Transfer, 115 (1993), 788. [5] G. Wang, P. Vanka, Int. J. Heat Mass Transfer, 38 (17) (1995), 3219. [6] T.A. Rush, T.A. Newell, A.M. Jacobi, Int, J. Heat Mass Transfer, 42 (1999), 1541.

272 13 International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, TR0700494

CRITICAL HEAT FLUX NEAR THE CRITICAL PRESSURE IN HEATER ROD BUNDLE COOLED BY R-134A FLUID: EFFECTS OF UNHEATED RODS AND SPACER GRID

Se-Young Chun*, Chan-Whan Shin, Sung-Deok Hong and Sang-Ki Moon Korea Atomic Energy Research Institute 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 Republic of Korea * E-mail: [email protected]

ABSTRACT A supercritical-pressure light water reactor (SCWR) is currently investigated as the next generation nuclear reactors. The SCWR, which is operated above the thermodynamic critical point of water (647 K, 22.1 MPa), have advantages over conventional light water reactors in terms of thermal efficiency as well as in compactness and simplicity. Many experimental studies have been performed on heat transfer in the boiler tubes of supercritical fossil fire power plants (FPPs). However, the thermal-hydraulic conditions of the SCWR core are different from those of the FPP boiler. In the SCWR core, the heat transfer to the cooling water occurs on the outside surface of fuel rods in rod bundle with spacers. In addition, the experimental studies in which the critical heat flux (CHF) has been carefully measured near the critical pressure have never yet been carried out, as far as we know. Therefore, we have recently conducted the CHF experiments with a vertical 5x5 heater rod bundle cooled by R- 134a fluid. The purpose of this work is to find out some novel knowledge for the CHF near the critical pressure, based on more careful experiments. The outer diameter, heated length and rod pitch of the heater rods are 9.5, 2000 and 12.85 mm, respectively. The critical power has been measured in a range of the pressure of 2.474.03 MPa (the critical pressure of R-134a is 4.059 MPa), the mass flux 5Ö2000 kg/m2s, and the inlet subcooling 4084 kJ/kg. For the mass fluxes of not less than 550 kg/m2s, the critical power decreases monotonously up to the pressure of about 3.63.8 MPa with increasing pressure, and then fall sharply at about 3.83.9 MPa as if the values of the critical power converge on zero at the critical pressure. For the low mass fluxes of 50 to 250 kg/m2, the sharp decreasing trend of the critical power near the critical pressure is not observed. The CHF phenomenon near the critical pressure no longer leads to an inordinate increase in the heated wall temperature such as the case of the DNB (departure from nucleate boiling) at normal pressure conditions. The existence of a threshold pressure at which the CHF phenomenon disappears has been observed near the critical pressure. In the region where the pressure passes across the threshold pressure, CHF does not occur and the wall temperature variations increase monotonously according to the power level applied to the heater rods. The effects of unheated rods and spacer grid with mixing vane on the critical power have been investigated. The effect of unheated rods in the rod bundle on the critical power becomes smaller as the pressure approaches the critical pressure, and when the pressure exceeds 3.9 MPa, the unheated rods have little effect on the critical power. In the case of the rod bundle with the mixing vane spacer grids, the critical power shows larger value compared to that for the spacer grids without mixing vane. This trend is kept up to the pressure of 4.0 MPa (P/Pc =9.85) very close to the critical pressure.

273

PARALLEL SESSION 15C: LOW ENERGY REACTION AND NUCLEAR PHYSICS 13' International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, TR0700425

INVESTIGATION OF EXCITATION FUNCTIONS USING NEW EVALUATED EMPIRICAL AND SEMI-EMPIRICAL SYSTEMATIC FOR 14-15 MEV (N,T) REACTION CROSS SECTIONS

E. Tel1, A. Aydin3, E. G. Aydın1, A. Kaplan4 *Gazi University, Faculty of Art and Science, Ankara, Türkiye 3Kirikkale University, Faculty of Art and Science, Kirikkale, Türkiye 4 Süleyman Demirel University, Science and Letter Faculty, Department of Physics, İSPARTA- TÜRKİYE

E-mail:

ABSTRACT The hybrid reactor is a combination of the fusion and fission processes. In the fusion-fission hybrid reactor, tritium self-sufficiency must be maintained for a commercial power plant. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. Working out the systematics of (n,t) reaction cross sections is of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at energies up to 20 MeV. In this study, we have investigated the asymmetry term effect for the (n,t) reaction cross sections at 14-15 neutron incident energy. It has been discussed the odd- even effect and the pairing effect considering binding energy systematic of the nuclear shell model for the new experimental data and new cross section formulas (n,t) reactions developed by Tel et al. We have determined a different parameter groups by the classification of nuclei into even-even, even-odd and odd-even for (n,t) reactions cross sections. The obtained empirical formulas by fitting two parameter for (n,f) reactions were given. All calculated results have been compared with the experimental data. By using the new cross sections formulas (n,t) reactions the obtained results have been discussed and compared with the available experimental data. Keywords: Tritium, (nj) cross-section, empirical and semi empirical formulas, hybrid reactor, fusion reactor,

276 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Î TR0700426 LOW ENERGY NUCLEAR REACTIONS: 2007 UPDATE

Steven B. Krivit New Energy Times, 11664 National Blvd. #142, Los Angeles, CA 90064, USA E-mail: [email protected]

ABSTRACT Introduction This paper presents an overview of the field of low energy nuclear reactions (LENR), a branch of condensed matter nuclear science. It explains some of the various terminologies that have been used to describe this field since it debuted as "cold fusion" in 1989. The paper also reviews some of the most interesting news and developments regarding low energy nuclear reaction experiments and theory, and some of the sociological and political trends that have affected the field over the last 18 years. It concludes with a list of resources and information for scientists, journalists and decision makers.

Understanding the Nature of the Reactions

The worldwide LENR research effort includes 200 researchers in 13 nations. Over the last 18 years, 12 international conferences have been held, as well as 7 regional conferences in Italy, 14 in Russia and 7 in Japan.

The significant questions that face this field of research are: a) Are LENRs a genuine nuclear reaction? b) If so, is there a release of excess energy? and c) Are transmutations possible?

If the answers to these questions turn out to be positive, the next questions will be: d) Is the energy release cost-effective? and e) Are the transmutations useful?

Despite the fact that repeatability and reproducibility are challenging, the required parameters for achieving the excess heat effect are well understood. First, a high atomic loading ratio of D into Pd is required. In most conditions, 0.90 is the minimum threshold required to produce an excess heat effect. Second, a high electrical current density in the cathode is needed, 250 mA/cm2 under most conditions. The third requirement is for some kind of dynamic trigger to impose a deuterium flux in, on or around the cathode. The challenge that researchers face is how to achieve these conditions.

Some of the Most Interesting Research Developments

Work by Stanislaw Szpak, Pamela Boss and Frank Gordon at the U.S. Navy's SPA WAR Systems Center in San Diego has demonstrated the most remarkable evidence of nuclear reactions in the 18-year history of this field: evidence of charged particles in unprecedented magnitudes. A brief overview of their work is provided.

The published theory proposed by Allan Widom and Lewis Larsen will be briefly reviewed. No technical faults with the theory have been reported to date. The theory may be able to explain what have been described as the three miracles of "cold fusion": the lack of strong

277 13th International Conference on Emerging Nuclear Energy Systems June 03-08,2007, Istanbul, Türkiye neutron emissions; the mystery of how the Coulomb barrier is penetrated; and the lack of strong emission of gamma or x-rays.

Sociological and Political Trends

Public and scientific perception of LENR over the last 18 years has strongly affected progress in the field. This influence will be considered, as will the sociological and political impact of potentially paradigm-breaking science research and exploration in general.

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13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, 111 TR0700427

EXPERIMENTAL STUDIES ON COLD FUSION AND HYDROGEN - METAL

Docteur PEUBE-LOCOU Physique Ucleaire, Administrateur unique statutaire LES STRADETZ, 47340, LAROQUE-TIMBAUT, France E-mail: SCM-INTERNATIONALE(a),swissinfo.ors

ABSTRACT Cold fusion The cold fusion is a nuclear fusion realized in pervading conditions of temperature and pressure. My own process is parallel to that of the team of the University of Los Angeles, but shaped in 1996 within my personal and private Laboratory:

A small cylinder in which we replace the air by some deuterium to the gas state in - 33 ° (the deuterium is some hydrogen with a neutron, which we find in quantity in the sea water). We introduce a crystal there, extremely rare - the property of which is to emit continuously one thousand times dose of successful energy and it during several years without outside pyro, natural excitement - electric - that is it creates an electric field in the slightest change in temperature. We carry then the whole in + 7 °, what generates in some seconds a 200 000 volt electric field, an intense enough field to separate the pits of the deuterium of their electrons and to admit them to collide with those of the crystal. There is a fusion of protons between them (positive particles of the pit (core)) and a emission of neutrons, which do not merge. It is this emission which serves for measuring the quantity of energy produced by the fusion (merger). We so managed to produce some energy in unlimited quantity, allowing us to feed our installation with electric current in total autarky, and to reduce so our costs of functioning to divers domains. This crystal is exceptional in its applications and to give it the name would return has to break our current headway: the thorough problem, in this current period of takeover by the financial bodies of the possible patents, brought to us to the biggest caution as regards our results. And, as we look for no outside financing, we do not need to submit ourselves to the requirements lauded by the scientific Community, only our results are strictly estimated...

For example we can make estimate our bars or patches of Hydrogen - Metal: a simple spectroscopy in YEW will give as result, only, the element H. Indeed one of three geneses focusing for the obtaining of the Hydrogen - pure Metal with one thousand thousandth, and stable HTBT, uses this process of cold fusion (merger) above described. There is another process of Cold Fusion (Merger) which also gives, and in a more continuous way, the possibility to obtain from some Hydrogen - stable Metal and in bigger quantity by module of production:

When a metal as the scandium is used in fusions with cold, he (it) goes out of it covered with eruptions looking like microphones (micro computing) volcanoes. This puts in evidence the waste of a very big quantity of heat pulling (entailing) the cast iron and the evaporation of the metal, but only in tiny points. Knowing the melting point of the scandium, we calculated the quantity of energy necessary for the fusion (merger) of these nano - points without provoking of fusion (merger) surrounding. Seen the unhoped-for figure which we found, the source of energy can be only nuclear.

279 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, Istanbul, Türkiye

My spectrometer allows an exact measure of atoms and determined a broadcast (emission, issue) of helium (10 for 10.000). This got back heat allows the usage of an unlimited energy. What counts above all, is the conservation of Process and the acquisition of the results, their possible putting at disposition on the world plan(shot), under certain conditions, which (who) are not necessarily expensive.

Hydrogen - Metal

Everybody knows that the common hydrogen is a flammable gas of density very tenue. However, if we compress very strongly the hydrogen, the theory predicts that it is transformed into metal, and it would be possible that this metal phase is stable in common temperature. In fact, quite as the antimatter, the metal hydrogen constitutes a subject of researches mattering within the military laboratories for a very long time. One of the concrete reasons of this interest lives in the fact that the metal hydrogen is probably the most powerful chemical explosive that it is possible to conceive.

It is also the unique inexhaustible source of Energy for all the manners, appropriate , non- polluting, being able to be produced in unlimited quantity, in a cost of 1.000 USD the ton, once experience the modules of production the special patent of which we also have, put down(deposited) except community Europe in 1997.

We can within the Conference confide a bar of 5 grams to experts to determine the structure, and who will have to publish the survey report.

The synthesis of the metal hydrogen was not realized this day by the others laboratory yet although the theory indicates that such a synthesis will be possible with equipments such as the Laser Megajoule, according to the official thesis.

Nevertheless the SCM produced in 1987 already five patches of 5 grams of this new stable metal HTBT, in 1998 two bars of 1.000 grams - (Accompanying document 1)-, and in 2.001 of the Deuterium - Metal for a 50 gram bar.

It is a major discovery for the energy survival of the Humanity.

280 13th International Conference on Emerging Nuclear Energy Systems June 03-08, 2007, TR0700428

TUNNELING EFFECT ENHANCED BY LATTICE SCREENING AS MAIN COLD FUSION MECHANISM: AN BRIEF THEORETICAL OVERVIEW

Fulvio Frisone Department of Physics, University of Catania, Via Santa Sofia 64, 95123 Catania, Italy E-mail: [email protected]

ABSTRACT In this paper are illustrated the main features of tunneling traveling between two deuterons within a lattice. Considering the screening effect due lattice electrons we compare the d-d fusion rate evaluated from different authors assuming different screening efficiency and different d-d potentials. Then, we propose a effective potential which describe very well the attractive contribute due to plasmon exchange between two deuterons and by means of it we will compute the d-d fusion rates for different energy values. Finally the good agreement between theoretical and experimental results proves the reality of cold fusion phenomena and the reliability of our model.

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286 ISBN-978-975-01805-0-7 HEAD OFFICE Göreme Sokak No: 8/5, ICENES 2007 ORGANIZED BY 06680 Kavaklıdere - Ankara / TURKEY +90.312.466 15 00(pbx) +90.312.466 15 25 +90.312.466 12 95 [email protected] 8AKHUS TOURISM INDUSTRY & TRADING INC congress@ bakhus. com. tr [email protected] www.bakhus.com.tr