Symposium on Water Chemistry and Corrosion in Nuclear Power
Transcription
Symposium on Water Chemistry and Corrosion in Nuclear Power
Book of Abstracts Asian Water Chemistry Symposium Series AWC 2015 Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia - 2015 02 - 04 September, 2015 Convention Centre, Anupuram, India Organized by Indian Nuclear Society In association with Water and Steam Chemistry Division, BARC DAE Advisory Committee on Steam and Water Chemistry (COSWAC) Supported by Indian Society for Radiation and Photochemical Science (ISRAPS) Board of Research in Nuclear Sciences (BRNS) Atomic Energy Regulatory Board (AERB) Nuclear Power Corporation of India Limited (NPCIL) Association of Environmental Analytical Chemistry of India (AEACI) Foreword We are happy to host the Asian Water Chemistry Conference entitled ‘Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia-2015’ (AWC-2015) during September 02-04, 2015 at the Convention Centre Anupuram, Kalpakkam (Tamil Nadu). This is the first ever Asian Water Chemistry Conference being organized in India. The symposium is organized jointly by the Indian Nuclear Society (INS), Committee on Steam & Water Chemistry (COSWAC) of the Department of Atomic Energy (DAE) and the Water & Steam Chemistry Division (WSCD) of Bhabha Atomic Research Centre (BARC). DAE has attached critical importance to R&D in water chemistry and corrosion, and WSCD in close association with COSWAC is contributing to the high standards of water chemistry in Indian nuclear reactors in accordance with plant safety and regulatory requirements. The organizers of AWC-2015 are thankful to the Core Committee for giving them the responsibility of organizing this event in India, thereby providing the Indian researchers, nuclear plant chemists and operators an opportunity to showcase their R&D work and plant experience. AWC-2015 has received an overwhelming response from R&D institutes and utilities of India, Japan, South Korea, Taiwan, China and Germany. Indian participation is drawn from BARC, Nuclear Power Corporation of India, Atomic Energy Regulatory Board, Bharatiya Nabhikiya Vidyut Nigam, Indira Gandhi Centre for Atomic Research, Heavy Water Board, Institute of Plasma Research, Raja Ramanna Centre for Advance Technology, M/s Larson & Toubro, M/s Adani Infra India Ltd, and a few academic institutes. Over 150 delegates including 20 from overseas are expected to participate in the symposium. AWC-2015 will have 3 plenary talks, 7 invited talks, 23 technical talks and over 50 poster presentations covering topics of concern to PWR, BWR, PHWR, AHWR and VVER. The abstracts of these talks/presentations are compiled in this volume and they cover a wide spectrum of issues such as primary system operating experience, chemistry of boiling water reactor and other advanced reactor, corrosion of plant materials, decontamination and dose reduction, basic research in water chemistry steam cycle performance and process water, and issues relating to condenser cooling water. We are confident that AWC-2015 will provide a useful platform for deliberations on recent advances and thus focusing in the challenges and opportunities in the area of water chemistry and corrosion. We gratefully acknowledge the valuable advice of the Advisory Committee and the dedicated work of the members of the Symposium Organizing Committee and Local Organizing Committee. We extend our warm welcome to all the delegates from India and abroad, and wish them memorable and intellectually satisfying stay over three days of the symposium. Dr. B.N. Jagatap Chairman, Organizing Committee On behalf of the Organizing and Local Organizing Committees National Advisory committee Chairman Dr. R. K. Sinha, Secretary, DAE Members Shri. S. S. Bajaj, Chairman, AERB, Mumbai Shri. Sekhar Basu, Director, BARC, Mumbai Dr. P. R. Vasudeva Rao, Director, IGCAR, Kalpakkam Dr. P. Chellapandi, CMD, BHAVINI, Kalpakkam Dr. Vijayamohanan K. Pillai, Director, CECRI, Karaikudi Shri. T. J. Koteeswaran, Station Director, MAPS, Kalpakkam Dr. M. A. Atmanand, Director, NIOT, Chennai Dr. B. N. Jagatap, Director, Chemistry Group, BARC, Mumbai Dr. K. L. Ramakumar, Director, Isotope Group, BARC, Mumbai Dr. P. K. Vijayan, Director, RDDG, BARC, Mumbai Dr. J. K. Chakravartty, Director, Materials Group, BARC, Mumbai Shri. N. Nagaich, ED (CP&CC), NPCIL Shri Ravindranath, ED (LWR), NPCIL Shri U.C. Muktibodh, ED (Engg), NPCIL Dr. T. Jayakumar, Director, MMG, IGCAR, Kalpakkam Shri. Amitava Roy, Facility Director, BARC-F, Kalpakkam Dr. Usha Natesan, Director, Centre for Research, Anna University, Chennai Dr. J. B. Joshi, Homi Bhabha Chair, HBNI, Mumbai Shri. S. A. Bharadwaj, President, Indian Nuclear Society, Mumbai Dr. S. K. Apte, HBNI, Mumbai Dr. R. K. Singh, Secretary, Indian Nuclear Society, Mumbai Core Committee Representatives Prof. Yosuke Katsumura, Tokyo University, Japan Prof. Xinqiang Wu, Key Laboratory of Institute of Metal Research, Chinese Academy of Sciences, China Prof. Tsung Kuang Yeh, National Tsing Hua University, Taiwan Prof. In Hyoung Rhee, Soonchunhyang University, South Korea National Organizing committee Dr. B. N. Jagatap, BARC, Chairman Dr. D. K. Palit, ISRAPS, BARC Dr. S. Velmurugan, WSCD, Convener Dr. B. S. Panigrahi, FBTR, IGCAR Dr. S. K. Aggarwal, Retd. AD, RC&I group, BARC Prof. V. S. Raja, IIT-Mumbai Dr. C. M. Das, BRNS, BARC Shri. A. K. Rajput, RAPS#7&8, NPCIL Dr. S. Dutta, BARC Shri. R. Ramakrishnan, PRPD, BARCF Dr. K. Hari Krishna, MAPS, NPCIL Dr. S. Rangarajan, WSCD, BARC-F Shri. Y. V. Harinath, WSCD, BARC-F (Treasurer) Dr. A. L. Rufus, WSCD, BARC-F Smt. S. Jayashree, RED, BARC Shri. R. R. Sahaya, NPCIL Dr. U. Kamachi Mudali, IGCAR Dr. B. Sengupta, NPCIL Shri. T. V. Krishna Mohan, WSCD, BARC-F Dr. R. S. Sharma, ROMG, BARC Shri. Mahendra Prasad, AERB Dr. B. Venkatraman, INS, Kalpakkam Branch Dr. P. K. Mathur, Retd. Head, WSCL Dr. V. P. Venugopalan, WSCD, BARC-F Dr. D. B. Naik, RPCD, BARC Dr. G. K. Vithal, HWB Dr. S. V. Narasimhan, Retd. AD, CG, BARC Dr. Vivekanand Kain, MSD, BARC Prof. E. Natarajan, Anna University Symposium organisers would like to thank the following organisations for supporting the Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia, 2015 September 2 (Day-1) Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia–2015, Convention Centre, Anupuram, India 09:30 – 10:00 Opening Ceremony Inauguration 10:00 – 11:05 Plenary Session Plenary Lecture 1 10:00 – 10:30 Plenary Lecture 2 10:30 – 11:00 Summing up 11:00 – 11:05 Prof. Yosuke Katsumura, JRIA, Japan, AWC-102, “Current Situation of Nuclear Power in Japan after Fukushima Nuclear Accident” Shri. U. C. Muktibodh, NPCIL, India, AWC-160 “Chemistry and Corrosion Issues in Indian Nuclear Power Plants” Coffee Break/Photo Session 11:05 – 11:20 Session-1 Primary System Operating Experience (PWRs/PHWRs/VVERs) 11: 20– 12:30 Invited Talk 1 11:20 – 11:45 Technical Talk 1 11:45 – 12:05 Dr. S. V. Narasimhan, BARC (Retd.), India, AWC-167 “Chemistry Management in Indian Nuclear Reactors” Shri. P. Selvavinayagam, NPCIL, India, AWC-118 “Water Chemistry Experiences with VVER’s at Kudankulam” Dr. Hee-Sang Shim, KAERI , S. Korea, Technical Talk 2 12:05 – 12:25 AWC-150 “Prediction Method of Sub-Cooled Nucleate Boiling on the Nuclear Fuel Cladding in Primary Water Condition Using Acoustic Emission Technique” 12:25 – 12:30 Summing up Lunch Break Session-2 Boiling Water Reactor / Advanced Reactor Experience Invited Talk 2 13:30 – 13:55 12:30 – 13:30 13:30 – 15:00 Smt. Jayasree Sriram, BARC, India, AWC-157 “Water Chemistry Features of Advanced Heavy Water Reactor” Dr. Kenji Hisamune, JAPC, Japan, Technical Talk 3 13:55– 14:15 Technical Talk 4 14:15 – 14:35 Technical Talk 5 14:35 – 14:55 Summing up 14:55 – 15:00 Coffee Break AWC-101 “Approach to Mitigate Intergranular Stress Corrosion Cracking and Dose Rate Reduction by Water Chemistry Control in Tokai-2” Dr. Deepa Papachan, NPCIL, India AWC-133, “Fuel Performance Review at Taps 1&2 With Respect to Reactor Coolant Crud & Alpha Activity” Dr. Subrata Bera, AERB, India. AWC-185, “Iodine Chemistry and Associated Interaction Under Severe Accident Conditions” 15:00 – 15:15 15:15 – 17:05 Session-3 Corrosion of Plant Materials Invited Talk 3 15:15 – 15:40 Technical Talk 6 15:40– 16:00 Technical Talk 7 16:00– 16:20 Technical Talk 8 16:20 – 16:40 Dr. Vivekanand Kain, BARC, India AWC-169, “Correlating Size of Scallops to Single Phase Flow Accelerated Corrosion in Nuclear Power Plants” Dr. Helmut Nopper, AREVA, Germany, AWC-148, “ FAC Surveillance Concept with the Predictive Code COMSY” Prof. Yutaka Watanabe, Tohoku Univ., Japan, AWC-128, “Benchmark Study of Prediction Models for Pipe Wall Wastage” Shri. Mahendra Prasad, AERB, India, AWC-112, “ Mass Transfer Coefficient Enhancement Factor in Pipe Bend – 3 Dimensional Analysis” Dr. Yasuhiro Chimi, JAEA, Japan, Technical Talk 9 16:40 – 17:00 Summing up 17:00 – 17:05 AWC-113, “Water Radiolysis Effect on IASCC Growth Behavior in BWR Water Conditions in Highly Irradiated Austenitic Stainless Steel” 17:05 – 19:00 Poster Session Cultural Program 19:00 - 20:00 Dinner 20:00 - 21:30 Convention Centre, Anupuram End of first day September 3 (Day-2) Session-4 Decontamination and Dose Reduction Invited Talk 4 09:00 – 09:25 Technical Talk 10 09:25 – 09:45 Technical Talk 11 09:45 - 10:05 Summing up 10:05 – 10:10 09:00 – 10:10 Dr. S. Velmurugan, BARC, India, AWC-166 “Dose Reduction and Decontamination in Indian Nuclear Power Plant” Shri. Osamu Shibasaki, Toshiba Corporation, Japan, AWC-115 “Co-60 Deposition on Carbon-Steel Structural Materials after Seawater Infiltration in BWR Plant” Shri V. S. Sathyaseelan, BARC, India, Coffee Break AWC-162, “High Temperature Decontamination of Stainless Steel Surfaces” 10:10 – 10:25 10:25 – 11:35 Session-5 Basic Research Prof. Xinqiang Wu, IMR, China, Invited Talk 5 10:25 – 10:50 Technical Talk 12 10:50 – 11:10 Technical Talk 13 11:10 – 11:30 Summing up 11:30 – 11:35 AWC-100 "Development Status of Nuclear Power in China and Some Fundamental Research Progress on PWR Primary Water Chemistry in China" Dr. Puspalatha Rajesh, BARC, India, AWC-161, “Effect of High Concentration Gadolinium Nitrate in Reactor Moderator System” Dr. P. Chandramohan, BARC, India, AWC-136, “Cation Distribution in Ferrites and its Effects on The Chemical Dissolution Behaviour” Session-6 Steam Cycle Performance and Process Water Systems Invited Talk 6 11:35 – 12:00 Technical Talk 14 12:00 – 12:20 11:35-13:05 Dr. Christoph Michael Stiepani, AREVA, Germany, AWC-130,“AREVA’s Toolbox for Long-Term Best Performance and Reliable Operation of Nuclear Steam Generators” Dr. K. Ganapathy Subramanian, IGCAR, India, AWC -140, “Two Decades of Experience with Steam-Water Chemistry Maintenance of Fast Breeder Test Reactor” Dr. P. K. Pal, NPCIL, India, Technical Talk 15 12:20 – 12:40 Technical Talk 16 12:40 – 13:00 Summing up 13:00 – 13:05 AWC-105, “Identification of Boiler Tube Leak in RAPS-2 by Measuring Iodine-134 Activity in Boiler Water Sample of RAPS Using Gamma Spectrometric Techniques.” Dr. Meng Jen Chen, Taiwan power company Taiwan, AWC-170, “Improve Steam Generator Moisture Carryover Rate at Maanshan NPS by Cleaning Steam Drum Internal Sludge ” Lunch Break 13:05 – 14:00 Session-7 Condenser Cooling Water Issues 14:00 – 15:30 Invited Talk 7 14:00 – 14:25 Technical Talk 17 14:25 – 14:45 Technical Talk 18 14:45 – 15:05 Dr. V. P. Venugopalan, BARC, India, AWC-147, “Biofouling Control in Power Plant Cooling Water Systems: Challenges and Prospects” Dr.Vinita Vishwakarma, Sathyabama University, India, AWC-149, “Nanophase Modified Fly Ash Concrete with Superior Concrete Properties, Durability and Biofouling Resistance for Seawater Applications ” Dr. B. Anandkumar, IGCAR, India, AWC-151 “Control of Biofouling on Titanium Condenser Tubes with The Use of Electroless Copper Plating” Shri. A. K. Mohanty, IGCAR, India Technical Talk 19 15:05 – 15:25 Summing up 15:25 – 15:30 AWC-155 “Biofouling Community Pattern on Various Metallic Surfaces in The Coastal Waters of Kalpakkam, Southwestern Bay of Bengal ” Coffee Break 15:30 – 15:45 Session-8 Future Trends and New Developments 15:45 – 16:55 Invited Talk 8 15:45 – 16:10 Technical Talk 20 16:10 – 16:30 Technical Talk 21 16:30 – 16:50 Summing up 16:50 – 16:55 Dr. Alphonsa Joseph, IPR, India, AWC-109 “Plasma Nitrocarburizing Process –A Solution to Improve Wear and Corrosion Resistance” Dr. B. Anupkumar, BARC, India, AWC-173 “Metal Ion Imprinted Polymers for Effective Radioactive Waste Segregation in Nuclear Industry” Shri. Jaymin Gandhi, Adani Infra India Ltd., India, AWC-121, “Environmental Sustainability by Adoption of Alternate Cooling Media for Condenser Cooling” Coffee Break 16:55 – 17:10 Feedback Session 17:10 – 18:00 Dinner 19:00-21:00 NPCIL Guest House, Kalpakkam End of Second day September 4 (Day-3) Technical Visit 9:30:12:00 Visit to Madras Atomic Power Station and NDDP, Kalpakkam Lunch 13:00 :14:00 Convention Centre, Anupuram End of Technical program Visit to ‘DakshinaChitra’ 14:00:18:00 ‘DakshinaChitra’ is an exciting cross cultural living museum of art, architecture, lifestyles, crafts and performing arts of South India. (www.dakshinachitra.net) S. No. UID Title and Authors P.No Plenary Session 1 AWC-102 2 AWC100 3 AWC-160 CURRENT SITUATION OF NUCLEAR POWER IN JAPAN AFTER FUKUSHIMA NUCLEAR ACCIDENT Yosuke Katsumura DEVELOPMENT STATUS OF NUCLEAR POWER IN CHINA AND FUNDAMENTAL RESEARCH PROGRESS ON PWR PRIMARY WATER CHEMISTRY IN CHINA Xinqiang Wu, Xiahe Liu, En-Hou Han, Wei Ke, and Yuming Xu INDIAN NUCLEAR POWER PRORAMME AND EXPERIENCE IN WATER CHEMISTRY AND CORROSION U.C Mukthibodh 1 2 - Session –I 1 AWC-167 CHEMISTRY MANAGEMENT IN INDIAN NUCLEAR REACTORS S.V. Narasimhan 3 2 AWC-118 WATER CHEMISTRY EXPERIENCES WITH VVERs AT KUDANKULAM D. Rout, T.C. Upadhyay, Ravindranath, P. Selvinayagam and R.S. Sundar 4 AWC-150 PREDICTION METHOD OF SUB-COOLED NUCLEATE BOILING ON THE NUCLEAR FUEL CLADDING IN PRIMARY WATER CONDITION USING ACOUSTIC EMISSION TECHNIQUE Hee-Sang Shim, Seung-Heon Baek, Kaige Wu, Deok Hyun Lee and Do Haeng Hur 5 3 Session -II AWC-157 WATER CHEMISTRY FEATURES OF ADVANCED HEAVY WATER REACTOR Jayasree Sriram, Vivekanad Kain, S.Velmurugan and K.Vijayan 6 2 AWC-101 APPROACH TO MITIGATE INTERGRANULAR STRESS CORROSION CRACKING AND DOSE RATE REDUCTION BY WATER CHEMISTRY CONTROL IN TOKAI-2 Kenji Hisamune 7 3 AWC-133 FUEL PERFORMANCE REVIEW AT TAPS 1&2 WITH RESPECT TO REACTOR COOLANT CRUD & ALPHA ACTIVITY Deepa Papachan, A.K.Panda and S.M.Maskey 8 AWC-185 IODINE CHEMISTRY AND ASSOCIATED INTERACTION UNDER SEVERE ACCIDENT CONDITIONS Dhanesh B. Nagrale, Subrata Bera, Anuj Kumar Deo, U. K. Paul, M. Prasad and A. J. Gaikwad 9 1 4 S. No. UID Title and Authors P.No Session –III 1 AWC-169 2 AWC-148 3 AWC-128 4 AWC-112 5 AWC-113 CORRELATING SIZE OF SCALLOPS TO SINGLE PHASE FLOW ACCELEARATED CORROSION IN NUCLEAR POWER PLANTS Vivekanand Kain, V. Dubey, S. Roychowdhury, M. Kiran Kumar and D. K.Barua FAC SURVEILLANCE CONCEPT WITH THE PREDICTIVE CODE COMSY Helmut Nopper and Andre Zander BENCHMARK STUDY OF PREDICTION MODELS FOR PIPE WALL WASTAGE Yutaka Watanabe MASS TRANSFER COEFFICIENT ENHANCEMENT FACTOR IN PIPE BEND – 3 DIMENSIONAL ANALYSIS Mahendra Prasad, P Madasamy, T V Krishnamohan, S Velumurugan, Arunkumar Sridharan and Avinash J. Gaikwad WATER RADIOLYSIS EFFECT ON IASCC GROWTH BEHAVIOR IN BWR WATER CONDITIONS IN HIGHLY IRRADIATED AUSTENITIC STAINLESS STEEL Yasuhiro Chimi, Shigeki Kasahara, Kuniki Hata, Yutaka Nishiyama, Hitoshi Seto, Kazuhiro Chatani, Yuji Kitsunai and Masato Koshiishi 10 11 12 13 14 Session –IV 1 2 AWC-166 DEVELOPMENT OF METHODS TO CONTROL RADIATION FIELD AND CORROSION IN PHWRS S.Velmurugan 15 AWC-114 DEVELOPMENT OF A METHOD TO LOWER RECONTAMINATION AFTER CHEMICAL DECONTAMINATION BY DEPOSITING Pt NANO PARTICLES Tsuyoshi Ito, Hideyuki Hosokawa, Toshimasa Ohashi, Makoto Nagase,Mizuho Tsuyuki, Nobuyuki Ota and Motohiro Aizawa 16 3 AWC-115 4 AWC-162 Co-60 DEPOSITION ON CARBON-STEEL STRUCTURAL MATERIALS AFTER SEAWATER INFILTRATION IN BWR PLANT Hiromitsu Inagaki, Osamu Shibasaki, Koji Negishi, Yumi Yaita, Masato Okamura, Yutaka Uruma, Seiji Yamamoto and Hajime Hirasawa HIGH TEMPERATURE DECONTAMINATION OF STAINLESS STEEL SURFACES V. S. Sathyaseelan, A. L. Rufus, P. Chandramohan, H. Subramanian and S. Velmurugan 17 18 Session –V 1 AWC-161 EFFECT OF HIGH CONCENTRATION GADOLINIUM NITRATE IN REACTOR MODERATOR SYSTEM Debasis Mal, Puspalata Rajesh, S. Rangarajan and S. Velmurugan 19 2 AWC-173 METAL ION IMPRINTED POLYMERS FOR EFFECTIVE RADIOACTIVE WASTE SEGREGATION IN NUCLEAR INDUSTRY B. Anupkumar 20 3 AWC-136 CATION DISTRIBUTION IN FERRITES AND ITS EFFECTS ON THE CHEMICAL DISSOLUTION BEHAVIOUR P.Chandramohan, M.P.Srinivasan and S.Velmurugan 21 S. No. UID Title and Authors P.No Session -VI 1 AWC-130 AREVA’s TOOLBOX FOR LONG-TERM BEST PERFORMANCE AND RELIABLE OPERATION OF NUCLEAR STEAM GENERATORS Andreas Drexler, Steffen Weiss, Neil Caris and Christoph Stiepani 22 2 AWC-140 TWO DECADES OF EXPERIENCE WITH STEAM-WATER CHEMISTRY MAINTENANCE OF FAST BREEDER TEST REACTOR K. Ganapathy Subramanian, A. Suriyanarayanan and B. S. Panigrahi 23 3 AWC-105 4 AWC-170 IDENTIFICATION OF BOILER TUBE LEAK IN RAPS-2 BY MEASURING IODINE-134 ACTIVITY IN BOILER WATER SAMPLE OF RAPS USING OF GAMMA SPECTROMETRIC TECHNIQUES. P. K. Pal and R. C. Bohra IMPROVE STEAM GENERATOR MOISTURE CARRYOVER RATE AT MAANSHAN NPS BY CLEANING STEAM DRUM INTERNAL SLUDGE Meng-Jen Chen 24 25 Session -VII 1 AWC-147 2 AWC-149 3 AWC-151 4 AWC-155 BIOFOULING CONTROL IN POWER PLANT COOLING WATER SYSTEMS: CHALLENGES AND PROSPECTS V.P.Venugopalan NANOPHASE MODIFIED FLY ASH CONCRETE WITH SUPERIOR CONCRETE PROPERTIES, DURABILITY AND BIOFOULING RESISTANCE FOR SEAWATER APPLICATIONS Vinita Vishwakarma, R.P. George U. Sudha, D. Ramachandran, Kalpana Kumari, R. Preetha, U.Kamachi Mudali and C. S. Pillai CONTROL OF BIOFOULING ON TITANIUM CONDENSER TUBES WITH THE USE OF ELECTROLESS COPPER PLATING B. Anandkumar, R.P. George, D. Ramachandran and U. Kamachi Mudali BIOFOULING COMMUNITY PATTERN ON VARIOUS METALLIC SURFACES IN THE COASTAL WATERS OF KALPAKKAM, SOUTHWESTERN BAY OF BENGAL Gouri Sahu, K.K. Satpathy, A.K.Mohanty and V.K.Bindu 26 27 28 29 Session –VIII 1 AWC-109 PLASMA NITROCARBURIZING PROCESS –A SOLUTION TO IMPROVE WEAR AND CORROSION RESISTANCE Alphonsa Joseph, Ghanshyam Jhala and S. Mukherjee 30 2 AWC-154 FEASIBILITY STUDY OF A NON-CHEMICAL TECHNIQUE FOR FOULING CONTROL Yaw-Ming Chen 31 3 AWC-121 ENVIRONMENTAL SUSTAINABILITY BY ADOPTION OF ALTERNATE COOLING MEDIA FOR CONDENSER COOLING Jaymin Gandhi and Nilesh Patel 32 S. No. UID Title and Authors P. No Poster Session AWC-103 RADIOLYSIS OF WATER AT HIGH TEMPERATURE AND PRESSURE CONDITIONS: A PICOSECOND PULSE RADIOLYSIS EXPERIMENT AND NUMERICAL SIMULATIONS Yusa Muroya, Tetsuro Yoshida, Yosuke Katsumura, Shinichi Yamashita, Mingzhang Lin and Takahiro Kozawa 33 AWC-104 INJECTION OF NANO-PARTICLES IN MITIGATING FLOW ACCELERATED CORROSION(FAC) DAMAGE IN THE SECONDARY SYSTEM OF NUCLEAR POWER PLANTS(NPPS) Dong Seok Lim, Hee Kwon Ku and Jae Seon Cho 34 AWC-106 ADDITION OF OXYGEN IN THE INLET OF RECOMBINER UNIT IN MODERATOR COVER GAS SYSTEM TO FACILITATE RECOMBINATION OF DEUTERIUM AND OXYGEN TO BRING DEUTERIUM CONCENTRATION IN SAFE LIMITS P. K. Pal and S. Mukherjee 35 4 AWC-107 DETERMINATION OF MOISTURE CONTENT IN STEAMS BY ANALYZING SODIUM CONTENT IN STEAM GENERATOR WATER & STEAMS CONDENSATE OF A NUCLEAR POWER PLANT USING ION CHROMATOGRAPHIC TECHNIQUE AT DIFFERENT LEVELS OF BOILER WATER P.K.Pal and R.C.Bohra 36 5 AWC-108 EXPERIENCE ON KKNPP VVER 1000 MWe WATER CHEMISTRY S. Ganesh, S. Selvaraj, M.R. Balasubramanian, P. Selvavinayagam#, Suresh Kumar Pillai, 37 6 AWC-110 KINETICS OF DISSOLUTION OF Ni-Cr CONTAINING IRON OXIDES SERIES (NICRXFE2-XO4) IN HMnO4 MEDIUM V. Balaji, P. Chandramohan, Ashish Tiwari, S. Rangarajan and S. Velmurugan 38 7 AWC-111 MEASUREMENT OF HENRY’S LAW CONSTANT IN HYDROGENATED LIOH/H3BO3 SOLUTION E. H. Lee, G. G. Lee, D. H. Lee, and D. H. Hur 39 AWC-116 SEASONAL VARIATION IN TRIHALOMETHANE LEVELS AT KALPAKKAM AND IN RELATION TO ORGANIC CARBON PRECURSORS R. Rajamohan, V. P. Venugopalan and Usha Natesan 40 1 2 3 8 9 AWC-117 10 AWC-119 11 AWC-120 12 AWC-122 ROLE OF REDUCTANTS IN CONTROL OF CORROSION OF MATERIALS RELATED TO NUCLEAR REACTORS Padma S.Kumar, Sinu Chandran, Puspalata Rajesh, D.Mohan, S.Rangarajan and S.Velmurugan IMPROVEMENT IN PERFORMANCE OF DM PLANT, SECONDARY SYSTEMS FOR ACHIEVING CHEMISTRY PERFORMANCE INDICATOR OF KGS-3&4 B.S.Sahu, P.G.Raichur, M Srinivas and M P Hansora EXPERIENCE OF CHEMICAL TREATMENT FOR CONTROLLING CORROSION IN IDCT WATER OF KGS 3&4. V.Uday Kumar and B.S.Sahu ELECTROCHEMICAL PASSIVATION STUDIES OF ZIRCALOY IN PRESENCE OF METAL ION Sinu Chandran, H. Subramanian, N. Sreevidya, S. Rangarajan and S. Velmurugan 41 42 43 44 S. No. UID Title and Authors P.No Poster Session A COMPARATIVE STUDY OF THE CORROSION BEHAVIOUR OF GRADE 91 AND RAFM STEELS AT AMBIENT TEMPERATURE N Sreevidya, Sinu Chandran, C.R. Das, S. K Albert, S Rangarajan and S Velmurugan PERFORMANCE RESTORATION TECHNIQUE DEVELOPED FOR FOULED HEAT EXCHANGER Dipankar Nanda, Babloo Tiwari and R. M. Pandey NITROGEN COMPOUNDS FORMATION IN N2-WATER AND N2MOISTURE SYSTEMS G.R. Dey and T.N. Das EVALUATION OF ALUMINUM BRASS COUPONS IN BWR CONDENSATE ENVIRONMENT IN PRESENCE OF METAL K K Bairwa, V S Tripathi, A Kumar and D B NaiK 13 AWC-123 14 AWC-124 15 AWC-125 16 AWC-126 17 AWC-127 SYNTHESIS AND CHARACTERIZATION OF V(HCOO)2•2H2O V S Tripathi K K Bairwa1, S N Achary and D B. NaiK 49 18 AWC-129 STUDIES ON FAILURE ANALYSIS OF STAINLESS STEEL ION EXCHANGE HOPPER AT NAPS S.K.Upadhyay, Ranjana Kusari, and Brij Mohan 50 19 AWC-131 20 AWC-132 21 AWC-134 22 AWC-135 23 AWC-137 24 AWC-138 25 AWC-139 ANTIMONY (Sb) SORPTION AT HIGH TEMPERATURE AND PRESSURE ON ZIRCALOY, CARBON STEEL (CS) AND MAGNETITE COATED CS (MCS) SURFACES S. J. Keny, B. K. Gokhale, A. G. Kumbhar, Santanu Bera, Saibal Basu and S. Velmurugan EFFECT OF ANTIMONY(III) ON CARBON STEEL CORROSION INHIBITION BY MOLYBDATE IN CITRIC ACID SOLUTION Vinit K. Mittal, Y. Raghavendra, Santanu Bera, S. Sumathi, S. Rangarajan, S.V. Narasimhan and S.Velmurugan RADIOACTIVE LIQUID WASTE DISCHARGE REDUCTION STRATEGIES AT TAPS 1&2 Deepa Papachan, A.K.Panda, S.M.Maskey, M.Joshi, and V.S.Daniel EVALUATION OF ADVANCED HOT CONDITIONING PROCESS FOR PHWRS P.Chandramohan, M.P.Srinivasan and S.Velmurugan TREATMENT OF FAST REACTOR LIQUID WASTEELECTROCHEMICAL METHOD Swapan Kumar Mahato, R. Sudha, P. Muralidaran and S. Anthonysamy FIXATION OF NUCLEAR WASTE INTO GLASS MATRICES FOR ULTIMATE DISPOSAL G. Hazra, T Das and P. Mitra ANTIMONY SORPTION PROPERTIES OF CHITOSAN – NANO TIO2 COMPOSITE BEADS Padala Abdul Nishad, B. Anupkumar and S. Velmurugan 26 AWC-141 HEAVY METALS-BIOREMEDIATION BY HIGHLY RADIORESISTANT DEINOCOCCUS RADIODURANS BIOFILM : PROSPECTIVE USE IN NUCLEAR REACTOR DECONTAMINATION Sudhir K. Shukla and T. Subba Rao 27 AWC-142 OPERATING CONDITIONS INFLUENCE CORROSION OF CARBON STEEL IN A FRESHWATER DISTRIBUTION SYSTEM T. Subba Rao 45 46 47 48 51 52 53 54 55 56 57 58 59 S. No. UID Title and Authors P.No Poster Session 28 AWC-143 ISOLATION AND CHARACTERIZATION OF THE MICROBIAL COMMUNITY OF A FRESHWATER DISTRIBUTION SYSTEM P. Balamurugan and T. Subba Rao 60 29 AWC-144 MICROFOULING ASSESSMENT AND ITS CONTROL IN A HEAVY WATER PRODUCTION UNIT Rajesh Kumar and T. Subba Rao 61 30 AWC-145 CORROSION OF ALLOY D9 IN LIQUID SODIUM R. Sudha, K. Chandran, P. Muralidaran and S. Anthonysamy 62 31 AWC-146 32 AWC-156 THREE DECADES OF EXPERIENCE WITH COOLING WATER SYSTEM OF A FAST REACTOR A.Suriyanarayanan and B.S.Panigrahi WATER TREATMENT WITH CHLORINE: INFLUENCE OF SOURCE WATER CHARACTERISTICS ON CHLORINATION & CBPS FORMATION R K Padhi, S Subramanian and K K Satpathy 63 64 33 AWC-158 ENTRAINMENT AND IMPINGEMENT OF AQUATIC FAUNA AT COOLING WATER SYSTEM OF MADRAS ATOMIC POWER STATION (MAPS) S. Barath Kumar, N. P. I. Das and K.K. Satpathy 34 AWC-159 SURFACE AND ELECTROCHEMICAL CHARACTERIZATION OF NANO ZINC FERRITE COATING ON CARBON STEEL Sumathi Suresh, S. Rangarajan and S velmurugan 35 AWC-163 36 AWC-171 37 AWC-172 DISSOLUTION OF COBALT METAL POWDER V.S.Sathyaseelan, A.L.Rufus and S.Velmurugan 69 AWC-174 STUDIES WITH ANTI FOULING COATING ON SEAWATER INTAKE SYSTEM SCREENS OF MAPS N.Sankar, V.S.Santhanam, P.Umapathi, K.Hari Krishna, D.Rajendran, P.S.Murthy and V.P. Venugopalan, 70 AWC-175 INFLUENCE OF GEOMETRY OF PIPE ON FLOW ACCELERATED CORROSION - A STUDY UNDER NEUTRAL PH CONDITION P.Madasamy, M.Mukunthan, P.Chandramohan, T.V.Krishna Mohan, and S.Velmurugan 71 AWC-176 EVALUATION OF PLASMA COATED CARBON STEEL TO RESIST FLOW ACCELERATED CORROSION P.Madasamy, J. Alphonsaa,J. Ghanshyam, S. Mukherjee, M.Mukunthan, P.Chandramohan, T.V.KrishnaMohan, ,E.Natarajan and S.Velmurugan 72 38 39 40 EVALUATION OF CORROSION INHIBITORS FOR HIGH TEMPERATURE DECONTAMINATION APPLICATIONS V. S. Sathyaseelan, A. L. Rufus and S. Velmurugan DEVELOPMENT OF LEACHING METHOD FOR THE ANALYSIS OF PALLADIUM CATALYST USED IN THE MODERATOR COVER GAS CIRCUIT OF MAPS BY ICP-OES S.Vijayalakshmi and S.Annapoorani 65 66 67 68 S. No. UID Title and Authors P. No Poster Session 41 AWC-177 PREPARATION AND DISSOLUTION OF URANIUM DIBUTYL PHOSPHATE (U-DBP) M.K.Dhanesh, A.L.Rufus and S.Velmurugan 73 42 AWC-178 STUDIES ON GADOLINIUM PRECIPITATION IN MODERATOR SYSTEM OF NUCLEAR REACTOR Akhilesh C Joshi, Puspalata Rajesh, A.L.Rufus and S.Velmurugan 74 OBSERVATIONS ON THE REMOVAL OF GADOLINIUM FROM THE MODERATOR SYSTEM OF PRESSURISED HEAVY WATER REACTOR (PHWR) AND ADVANCED HEAVY WATER REACTOR (AHWR) V. Praveena, Padma S.Kumar, A.L.Rufus and S.Velmurugan CHEMISTRY MANAGEMENT OF GENERATOR STATOR WATER SYSTEM N. Sankar, V.S. Santhanam, S.R. Ayyar, P. Umapathi, P. Jeena, K. Hari Krishna, D.Rajendran STUDIES WITH SOLID CHLORINE CHEMICAL FOR CHLORINATION OF SEA WATER SYSTEMS N.Sankar, P.Kumaraswamy, V.S.Santhanam, P.Jeena, K.Hari Krishna and D.Rajendran, CORROSION RATE OF CARBON STEEL IN NEUTRON SHIELD TANK WATER R.Ramakrishnan, N.Rathinasamy and K.V.Ravi 43 AWC-179 44 AWC-180 45 AWC-181 46 AWC-182 47 AWC-183 OPTIMUM THICKNESS EVALUATION OF ZrO2 COATING ON TYPE 304L STAINLESS STEEL FOR CORROSION PROTECTION Nidhi Garg, Santanu Bera, V. S. Tripathi, Vijay Karki and S. Velmurugan 79 48 AWC-184 IODINE REMOVAL IN CONTAINMENT FILTERED VENTING SYSTEM DURING NUCLEAR ACCIDENT SubrataBera, D. B. Nagrale, Anuj Kumar Deo, U. K. Paul, M. Prasad and A. J. Gaikwad 80 49 AWC-186 AN OPERATIONAL EXPERIENCE WITH COOLING TOWER WATER SYSTEM IN CHILLING PLANT Manju B Rajan, Ankan Roy and KV Ravi 81 AWC-187 CONTAINMENT BEHAVIOR DURING MOLTEN CORIUM CONCRETE INTERACTION Anuj Kumar Deo, S. P. Lakshmanan, S. Bera, Balbir K. Singh, P. K. Baburajan, R. S. Rao, U. K. Paul and A. J. Gaikwad 82 AWC-188 DEUTERISATION OF MIXED BED ION EXCHANGE RESIN: KINETICS STUDY Satinath Ghosh, M. K. Tripathy, Kajal Dhole, T. Vasudevan, Satyam Shukla and R. S. Sharma 83 AWC-189 FEASIBILITY STUDY ON NANO-STRUCTURED COATINGS TO MIGATE FLOW-ACCELERATED CORROSION OF CARBON STEEL PIPING SYSTEM Seunghyun Kim, Jeong Won Kim and Ji Hyun Kim 84 50 51 52 75 76 77 78 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-102 CURRENT SITUATION OF NUCLEAR POWER IN JAPAN AFTER FUKUSHIMA NUCLEAR ACCIDENT Yosuke Katsumura Japan Radioisotope Association 2-28-35 Honkomagome, Bunkyo-ku, Tokyo 113-8941 Japan *Corresponding author: [email protected] ABSTRACT On March 11, 2011 we had the Great East Japan Earthquake and induced tsunami, which attacked the Fukushima Daiichi Nuclear Power Station (NPS). After Fukushima Nuclear Accident, not only all the power reactors but also all research reactors are out of service over four years. At the beginning of August a power reactor, Sendai #1, Kyushu Electric Power Company, restarted the operation for the first time and commercial operation will be approved after the inspection of start-up at the end of August or the beginning of September. In this talk, I would like to present briefly the situation of the NPS before and after Fukushima Nuclear Accident, new regulatory requirements for LWR plants by the Nuclear Regulation Authority, and current status and future of the Fukushima Daiichi NPS. Keywords: Fukushima Daiichi Nuclear Accident, new requirements, Nuclear Regulatory Authority, Decommissioning of Fukushima NPS 1 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 DEVELOPMENT STATUS OF NUCLEAR POWER IN CHINA AND FUNDAMENTAL RESEARCH PROGRESS ON PWR PRIMARY WATER CHEMISTRY IN CHINA AWC-100 Xinqiang Wu 1*, Xiahe Liu 1, En-Hou Han 1, Wei Ke 1, Yuming Xu 2 1 Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, P.R. China (* Tel: +86-24-23841883; Fax: +86-24-23894149; E-mail: [email protected]) 2 China nuclear energy association, 12 Chegongzhuang Street, Xicheng District, Beijing 100037, P.R. China *Corresponding author: [email protected] Abstract China's non-fossil fuels are expected to reach 20% in primary energy ratio by 2030. It is urgent for China to speed up the development of nuclear power to increase energy supply, reduce gas emissions and optimize resource allocation. Chinese government slowed down the approval of new nuclear power plant (NPP) projects after Fukushima accident in 2011. At the end of 2012, the State Council approved the nuclear safety program and adjusted longterm nuclear power development plan (2011-2020), the new NPPs’ projects have been restarted. In June 2015, there are 23 operating units in mainland in China with total installed capacity of about 21.386 GWe; another 26 units are under construction with total installed capacity of 28.5 GWe. The main type of reactors in operation and under construction in China is pressurized water reactor (PWR), including the first AP1000 NPPs in the world (units 1 in Sanmen) and China self-developed Hualong one NPPs (units 5 and 6 in Fuqing). Currently, China's nuclear power development is facing historic opportunities and also a series of challenges. One of the most important is the safety and economy of nuclear power. The optimization of primary water chemistry is one of the most effective ways to minimize radiation field, mitigate material degradation and maintain fuel performance in PWR NPPs, which is also a preferred path to achieve both safety and economy for operating NPPs. In recent years, an increased attention has been paid to fundamental research and engineering application of PWR primary water chemistry in China. The present talk mainly consists of four parts: (1) development status of China's nuclear power industry; (2) safety of nuclear power and operating water chemistry; (3) fundamental research progress on Zn-injected water chemistry in China; (4) summary and future. Keywords: China nuclear power, water chemistry, Zn-injection, fundamental research 2 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-167 CHEMISTRY MANAGEMENT IN INDIAN NUCLEAR REACTORS S.V. Narasimhan (ex) Chairman COSWAC, (Retd.) AD, CG, BARC *Corresponding author: [email protected] Abstract Starting with a couple of BWR, the department of atomic energy in India established PHWR based reactors of different capacities in quite a good number. Subsequently reactors of different types like VVER (already operational in Kudankulam), PWRs, AHWR, PFBR are being pursued vigorously to meet the energy demand and to comply with green power requirement. From the beginning of nuclear power programme, monitoring, maintaining & management of good chemistry domain in the cooling water circuits were practiced in a dedicated manner through the formation of an advisory group which in turn advises the national regulatory body on chemistry related issues. Enough care was taken to devise a suitable management model to implement these programmes in a meticulous manner. Over the years it has worked well not only in enabling the power plants to maintain water chemistry domain within the allowed specification limits but also in implementing newer techniques and procedures for enhancing operational safety and efficiency. The paper highlights some of these aspects. 3 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-118 WATER CHEMISTRY EXPERIENCES WITH VVERS AT KUDANKULAM 1 D Rout#, 1T.C. Upadhyay, 1Ravindranath, 2P. Selvinayagam & 2 R.S Sundar 1 Directorate of Operations, NPCIL, Mumbai 2 Kudankulam Nuclear Power Station, Kudankulam # Corresponding author: [email protected] Abstract Kudankulam Nuclear Power Project-1&2 (KKNPP-1&2) are pressurised water cooled VVERs of 1000 MW € each. KKNPP-1 is presently on its first cycle of operation and KKNPP-2 is on the advanced stage of commissioning with the successful completion of Hot Functional tests. Water Chemistry aspects during various phases of commissioning of KKNPP-1 such as Hot Run, Boric acid flushing, initial fuel Loading (IFL) , First approach to Criticality (FAC) are discussed. The main objectives of the use of controlled primary water chemistry programme during the hot functional tests are reviewed. The importance of the relevant water chemistry parameters were ensured to have the quality of the passive layer formed on the primary coolant system surfaces. The operational experiences during the 1st cycle of operation of primary water chemistry, radioactivity transport and build-up are presented. The operational experience of some VVER units in the field of the primary water chemistry, radioactivity transport and build-up are presented as a comparison to VVER at KKNPP. The effects of the initial passivated layer formed on metal surfaces during hot functional tests, activated corrosion products levels in the primary coolant under controlled water chemistry regime and the contamination/ radiation situation are discussed. This report also includes the water chemistry related issues of secondary water systems. 4 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-150 PREDICTION METHOD OF SUB-COOLED NUCLEATE BOILING ON THE NUCLEAR FUEL CLADDING IN PRIMARY WATER CONDITION USING ACOUSTIC EMISSION TECHNIQUE Hee-Sang Shim#, Seung-Heon Baek, Kaige Wu, Deok Hyun Lee, Do Haeng Hur Division of Nuclear Materials Safety Research, Korea Atomic Energy Research Institute (KAERI), Republic of Korea # Corresponding Author: [email protected] Abstract Axial offset anomaly (AOA), which is defined as a significant negative axial offset deviation from the predicted nuclear design value, has important operational and economic consequences. It is well known that AOA is caused by the incorporation of boron within corrosion product (crud) deposits on the upper span of fuel assembly. Crud depositions are accelerated when the sub-cooled nucleate boiling (SNB) occurs on the fuel cladding surface in the primary coolant including sufficient corrosion products. Many researchers have widely studied a boiling process via various methods such as high-speed video camera and temperature measurement with thermocouples to understand the crud deposition mechanism as well as SNB condition. The detection and monitoring of SNB in terms of non-destructive evaluation is one of promising technique for analyzing the crud deposition mechanism and AOA phenomenon. In this work, we provided a prediction method of SNB on fuel cladding using acoustic emission (AE) technique in a simulated primary circuit along with a relationship between crud deposition and boiling process. Crud deposition tests were performed in a simulated primary coolant including Ni- and FeEDTA of each 20 ppm at 325oC. The fuel cladding temperature was controlled using an internal heater in a temperature range of 330oC to 400oC and the boiling process was investigated using piezoelectric AE sensor coupled with fuel cladding surface. The transition of boiling process and bubble dynamics were successfully distinguished by AE signals in a primary coolant condition. In addition, the crud deposition depended on the boiling process in its properties and amount. These results indicate that the in-situ AE technique can be a suitable prediction method for SNB in the PWR reactor core. Keywords: Fuel cladding, Sub-cooled nucleate boiling, Acoustic emission, Crud deposition 5 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-157 WATER CHEMISTRY FEATURES OF ADVANCED HEAVY WATER REACTOR Jayasree Srirama#, Vivekanad kainb, S.Velmuruganc and K.Vijayana a Reactor Engineering Division, BARC, Mumbai b Materials Science Division, BARC, Mumbai c Water and Steam Chemistry Division, BARCF, kalpakkam # Corresponding Author: [email protected] Abstract Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various watercoolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS#2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. 6 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-101 APPROACH TO MITIGATE INTERGRANULAR STRESS CORROSION CRACKING AND DOSE RATE REDUCTION BY WATER CHEMISTRY CONTROL IN TOKAI-2 Kenji Hisamune The Japan Atomic Power Company 1-1, Kanda-Mitoshiro-Cho, Chiyoda-Ku, Tokyo *Corresponding author: [email protected] Abstract At the Tokai Daini Power Station (hereinafter Tokai-2; BWR, 1,100MWe, commenced commercial operation in November 1978), I carried out material replacement and stress release to maintain the integrity of structure materials. And, I reduced sulfate ion concentration by improvement of the regenerative method (such as the Advanced Resin Cleaning System; ARCS) of the condensate demineralizer ion-exchange resin to mitigate intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) materials. In addition, I suppressed reactor water oxide concentration by Hydrogen Water Chemistry during operation and start up to mitigate IGSCC. On the other hand, I worked on reduction of feed water iron concentration as the plant which I did not install a pre-filter in of condensate demineralizer for dose rate reduction. I improved operational change of condensate demineralizer ion-exchange resin regeneration and regenerative method (ARCS) for improvement of crud removal efficiency. In this report, I describe the improvement effect the water chemistry control (such as reduce of reactor water sulfate ion concentration, reactor water oxide concentration and feed water iron concentration) that I applied in Tokai-2 until now. In addition, I report dose rate reduction effect by the zinc injection that started an application recently. And, I introduce the ECP monitoring plan with the OLNCTM of the application plan in future. 7 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-133 FUEL PERFORMANCE REVIEW AT TAPS1&2 WITH RESPECT TO REACTOR COOLANT CRUD & ALPHA ACTIVITY Deepa Papachan#, A.K.Panda, S.M.Maskey Technical Services Section, Tarapur Atomic Power Station 1& 2 # Corresponding Author: [email protected] Abstract Tarapur Atomic Power Station -1&2 (TAPS-1&2) consists of twin unit of Boiling Water Reactors (BWR) in India. The reactors were commissioned during 1969-1970 and their present rated capacity is 530MWth. Each reactor core has 284 fuel assemblies.Fuel performance monitoring at TAPS1&2 is carried out throughout the fuel cycle by means of reactor physics parameters and several radiochemistry parameters e.g. gross gamma activity level of Iodine isotopes in reactor coolant, fission product noble gases(FPNG mainly Xe and Kr isotopes) radioactivity and their composition and monitoring of Cesium, TechniciumMolybdenum, Neptunium activity in reactor coolant. A comparative analysis of all these parameters helps in identifying the contribution of tramp uranium towards fuel performance, time of fuel failure occurrence and conjuring the extent of failed fuel growth etc. This feedback with regard to fuel performance during the fuel cycle not only guides in taking precautionary measures with respect to activites to be carried out at refueling floor during refueling outage, but it also increases the confidence level in leaky fuel detection at the end of the fuel cycle refueling outage by wet sipping method. Fuel Reliability Indicator(FRI) for BWR reactors, as provided by WANO is based on only the FPNG gases and is calculated for each reactor to monitor the industry trend at global level with respect to achieving high fuel integrity in spite of the fact that BWR reactors are of different generations. Along with the above mentioned fuel performance monitoring parameters and FRI, another radiochemical parameter which not only gives trend of fuel performance but also the clean up or filter system performance with respect to removal of failed fuel fragments or corrosion products is alpha activity of reactor coolant and fuel pool water. This paper compares the trend of primary coolant and fuel storage pool alpha activity with respect to the FRI trend of each reactor with an undernote that FRI reporting remains to be on higher side in spite of zero fuel failure owing to the insufficient deduction of tramp uranium factor from FRI calculation. 8 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-185 IODINE CHEMISTRY AND ASSOCIATED INTERACTION UNDER SEVERE ACCIDENT CONDITIONS Dhanesh B. Nagrale#, Subrata Bera, Anuj Kumar Deo, U. K. Paul, M. Prasad and A. J. Gaikwad Nuclear Safety Analysis Division, Atomic Energy Regulatory Board Niyamak Bhavan, Anushaktinagar, Mumbai-400094 # Corresponding author: [email protected] Abstract In a highly improbable severe accident wherein the core cooling is decapicitated or insufficient could lead to fuel elements melting and fission product release beyond the plant limits. Nuclear power plants are designed with engineering systems and associated operational procedures that provide an in-depth defence against such accidents. A good understanding of iodine behaviour is required for the analysis of severe accident consequences because Iodine is a major contributor to the potential source term for release to the environment. Iodine speciation along the transport path from fuel to cooler regions of heat transport system and into containment should be evaluated using various appropriate models which leads to prediction of volatile iodine mole fraction, cesium to iodine ratio etc. The roles of other elements mainly molybdenum, tellurium, uranium and lithium are also important. Iodine released from fuel, iodine transport in primary coolant system, reaction with control rods are important. The behaviour of iodine-bearing particles is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. Sorption and desorption of Iodine occurring on containment surface is also of importance. The presence of gaseous organic compounds, oxidizing compounds on iodine, reactions of aerosol Iodine with boron and formation of cesium iodide which results in more volatile Iodine release in containment are important aspects. Water radiolysis products due to presence of dissolved impurities such as dissolved oxygen, nitrate/nitrite (NO3 /NO2 ) produced by air radiolysis, trace metal ions such as Fe2+/Fe3+ dissolved from steel surfaces, chloride ions coming from the pyrolysis/radiolysis of polyvinyl material from cables and organic impurities (RH) from painted surfaces and polymers also important and all above mentioned phenomenon should be considered while calculating iodine released inside and outside containment. Other important aspects which needs attention are re-suspension from iodine loaded surfaces, coupling of thermal-hydraulics with iodine chemistry, temperature, relative humidity and steam condensation and its influence on the re-suspension rates, condensing conditions, etc. The released Iodine from containment is allowed to pass through containment filtration venting system (CFVS). CFVS consists of venturi scrubber and a scrubber tank. The scrubber tank which is dosed with NaOH, NaS2O3 where iodine will react with the chemicals and converts into NaI and Na2SO4. This paper elaborates about all above mentioned present status and issues with respect to iodine chemistry and its behaviour during accident. Keywords: Iodine, Volatile Iodine, Impurities, Organic and non-organic Iodine 9 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-169 CORRELATING SIZE OF SCALLOPS TO SINGLE PHASE FLOW ACCELEARATED CORROSION IN NUCLEAR POWER PLANTS Vivekanand Kain, V. Dubey, S. Roychowdhury, M. Kiran Kumar and D. K. Barua1 Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400085, India Nuclear Power Corporation of India Ltd., Anushaktinagar, Mumbai 400094, India # Corresponding author: Abstract A comprehensive thinning monitoring program is in place for all the components of high energy portions of secondary cycle systems of Indian nuclear power plants since 2006. This program is based on initially establishing a baseline thickness data by extensive ultrasonic examination on the components and then periodic thickness measurements to establish thinning and flow accelerated corrosion (FAC) rates. In this extensive thinning monitoring program, it has been observed that in a few cases the thinning are not by FAC degradation. This program has also identified components that are more prone to FAC and exhibit much higher FAC rates. These components are the focus of implementation of FAC thinning control measures e.g. change of material or change in piping layout/geometry to reduce flow rate/turbulence hence FAC. The extensive thickness monitoring data collected during FAC monitoring from various power plants provided an oppourtunity to correlate the size of scallops with the single phase FAC thinning rates. The extensive database thus developed indicates a broad trend. More specific measurements in different ranges of scallop sizes and FAC rates would be added to have a comprehensive database. Thus this provides an oppourtunity to establish FAC rate from the examination of the scallop pattern on the inner surface of the FAC affected component. This would be a very promising approach to confirm the FAC rate measured by ultrasonic examination. Specific case studies with emphasis on scallop patterns developed on cases of components showing low and high FAC rates are covered. A selective case would be highlighted to emphasise that factors other than FAC also lead to thinning. 10 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-148 FAC SURVEILLANCE CONCEPT WITH THE PREDICTIVE CODE COMSY Helmut Nopper*, Andre Zander AREVA GmbH Paul Gossen Str. 100, D-91058, Erlangen, Germany *Corresponding Author: [email protected] Abstract Flow-Accelerated Corrosion (FAC) has become a major safety issue for operators of NPPs. Experience shows, that severe wall thinning may cause spontaneously occurring pipe ruptures. It is therefore necessary to develop efficient surveillance concepts for wall thinning of piping, vessels and mechanical equipment. Wall tinning is typically caused by flow-induced degradation mechanisms like FAC, liquid droplet impingement erosion (LDI) and cavitation erosion (CE). Areas suffering from such wall thinning effects are difficult to locate, as these degradation mechanisms occur only locally for steel components operated under specific conditions of flow, water chemistry and temperatures. The reliable identification of system areas sensitive to wall thinning calls for a comprehensive plant-wide strategy, considering the possible superposition of different degradation mechanisms. The COMSY software supports this activity by providing the predictive capability to calculate degradation rates and the functionality to analyze the relevant operating conditions. Water chemical conditions are analyzed in each system and sub-system in respect to e.g. pH- values or oxygen concentrations and thermal hydraulic operation conditions are characterized for typical load cases experienced during plant service. Based on these evaluations a degradation potential can be quantified using predictive models. Based on these evaluations the inspection program can be optimized by focusing activities on degradation sensitive areas. An integrated inspection data management function ensures information feedback from inspection activities. Inspection data is systematically evaluated and used to further optimize service life predictions over the life cycle of the component. This strategy is designed to provide a comprehensive, long-term surveillance of the integrity of mechanical components. The computerized monitoring and lifetime surveillance system provided by AREVA makes it possible to keep a lifetime consumption record as a basis for safe operation and efficient maintenance and repair strategies. 11 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-128 BENCHMARK STUDY OF PREDICTION MODELS FOR PIPE WALL WASTAGE Yutaka Watanabe Tohoku University Aoba-ku, Sendai, Miyagi 980-8579, Japan # Corresponding Author: [email protected] Abstract The research committee for studying pipe-wall-thinning management was established in The Japan Society of Mechanical Engineers (JSEM) in Year 2008. Since then, the research committee has been gathering and investigating technical information on flow induced pipe wall wastage. As one of the core activities of the pipe-wall-thinning research committee, “working group for prediction methods” has been set up and prediction models of pipe-wall thinning have been reviewed in terms of their characteristics in prediction and required specification to be used in pipe-wall-thinning management. A few prediction models of flowaccelerated corrosion (FAC) and liquid droplet impingement erosion (LDI) have been reviewed by comparing the predictions to experimental data and plant inspection data. This paper describes the procedures and result status of the benchmark examinations. 12 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-112 MASS TRANSFER COEFFICIENT ENHANCEMENT FACTOR IN PIPE BEND – 3 DIMENSIONAL ANALYSIS Mahendra Prasad1*, P Madasamy2, T V Krishnamohan 2, S Velumurugan2, Arunkumar Sridharan3, Avinash J.Gaikwad1 1 Atomic Energy Regulatory Board, Mumbai, India 2 WSCD, BARCF, Kalpakkam, India 3 IIT Bombay, Mumbai, India # Corresponding author: [email protected] Abstract Flow Accelerated Corrosion (FAC) has plagued the power industry since long time. The high velocity fluid at elevated temperatures is used for process requirements which causes FAC in straight pipes exhibiting non-uniform corrosion and this is enhanced for junction such as bends, orifices etc. Mass transfer coefficient (MTC) changes from its base value in straight pipe (with same fluid parameters) for flow in bends, orifiec etc due to gross disturbance of the velocity profile. Since MTC is related to wall thinning, its relative increase or decrease is important to estimate for spatial degradation prediction. In this paper, computational fluid dynamics (CFD) simulations are carried for 58o bend angle and 2D bend radius circular pipe in three dimensions. Turbulent model K-ω with shear stress transport is found to perform well for domain with geometric changes and this was the model for simulation. Since mass tranfer boundary layer (MTBL) thickness δmtbl is related to the Schmidt number (Sc) and hydrodynamic boundary layer thicknessd δh, as δmtbl ~ δh/(Sc1/3), MTBL is significantly smaller than δh and hence boundary layer meshing was carried out deep into δmtbl. The simulation is related to an experiment carried out for 58o bend angle and 2D bend radius circular carbon steel pipe carrying water at 120oC under neutral pH conditions to determine the wall thinning at few extrados locations. Uniform velocity was applied at the inlet. The flow velocity was 5 m/s at room temperature. The ratio of the mass transfer coefficient at such locations to the straight pipe coefficient (MTCR) is determined through simulation. As seen in literature, since the dependence of MTCR on Re and Sc is not as strong as compared to pipe bend angle and bend radius, CFD simulation at lower temperature is sufficient to get approximate MTC in bends. The MTC increased in the extrados of the bend towards the outlet. Key Words: Flow Accelerated Corrosion, Mass Transfer Coefficient, Turbulent Flow, Mass Transfer Boundary Layer 13 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-113 WATER RADIOLYSIS EFFECT ON IASCC GROWTH BEHAVIOR IN BWR WATER CONDITIONS IN HIGHLY IRRADIATED AUSTENITIC STAINLESS STEEL Yasuhiro Chimi1,#, Shigeki Kasahara1, Kuniki Hata1, Yutaka Nishiyama1, Hitoshi Seto2, Kazuhiro Chatani2, Yuji Kitsunai2, Masato Koshiishi2 1 Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195, JAPAN 2 Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1313, JAPAN # Corresponding author: [email protected] Abstract For study of water radiolysis effect caused by gamma-rays from radioactive material on irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests in highly irradiated austenitic stainless steel are performed in simulated BWR water conditions (at ~563 K). The compact tension (CT) specimens made of 316L stainless steels are irradiated with neutrons up to ~12 dpa in the Japan Materials Testing Reactor (JMTR). Post-irradiation annealing at 973 K for 1 hour is applied to one of the specimens, which shows the recovery of material properties corresponding to the unirradiated ones but the radioactivity of highly irradiated material as it is. The gamma-ray absorbed dose rate in water is calculated near the crack tip of the CT specimen, and the stable concentrations of H2O2, O2 and H2 in water near the crack tip are estimated by radiolysis calculation for some feedwater conditions of normal water chemistry (NWC), deaerated water and hydrogen water chemistry (HWC). The preliminary results of the crack growth rate (CGR) for the highly irradiated specimens and the annealed specimen are presented, and the relationship between the CGRs and the water chemistry such as the concentrations of radiolytic species and the electrochemical corrosion potential (ECP) is discussed. 14 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-166 DEVELOPMENT OF METHODS TO CONTROL RADIATION FIELD AND CORROSION IN PHWRS S.Velmurugan Water and Steam Chemistry Division, BARC Facilities, Kalpakkam – 603 102 Tamilnadu, INDIA Corresponding author: [email protected] Abstract Pressurized Heavy Water Reactors (PHWRs) is the mainstay of Indian Nuclear Power Program. There are 18 PHWRs (220 MWe & 540 MWe) in operation and 4X700 MWe PHWRs are under construction. In these reactors, as far as radiation field is concerned, the philosophy of ALARA (As Low As Reasonably Achievable) is followed. The primary coolant system chemistry control is given due consideration during operation so that corrosion of structural material is minimized which in turn controls the radiation field. Development and application of full system Dilute Chemical Decontamination(DCD) process helped to reduce the radiation field in MAPS#1&2, RAPS#1&2, NAPS#1&2 and KAPS#1. PHWR being a tube type reactor, it enables application of full system decontamination to its heavy water primary coolant system. Significant reduction in radiation field and consequent savings in MANREM could be achieved. Attempts are being made to understand the problem created by the release of antimony activities (122Sb and 124Sb) during chemical decontamination and during planned shutdown. Passivation as a method to control the radiation field and corrosion is being studied. Magnesium ion as a passivator to the ferrite filmed structural materials of PHWRs is being investigated. In addition, as PHWRs uses carbon steel as structural material, the use of passivation as a method to control flow accelerated corrosion (FAC) is also being studied. Magnesium ion gets incorporated in the ferrite film formed over carbon steel structural material and is expected to reduce the solubility of magnetite film thereby the FAC of feeders in PHWRs. 15 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-114 DEVELOPMENT OF A METHOD TO LOWER RECONTAMINATION AFTER CHEMICAL DECONTAMINATION BY DEPOSITING PT NANO PARTICLES Tsuyoshi Ito1, Hideyuki Hosokawa1, Toshimasa Ohashi1, Makoto Nagase2, Mizuho Tsuyuki2, Nobuyuki Ota2, Motohiro Aizawa2 1 Center for Technology Innovation-Energy, Research & Development Group, Hitachi Ltd., 72-1 Omika-cho, Hitachi-shi, Ibaraki 319-1221 Japan 2 Hitachi Works, Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai-cho, Hitachi-shi, Ibaraki 3170073 Japan # Corresponding author: [email protected] Abstract Chemical decontamination is an effective method to reduce occupational radiation exposure in boiling water reactors (BWRs) when carrying out such large-scale tasks as overhauling primary recirculation pumps. In the chemical decontamination, oxides formed on the surface of the stainless steel (SS) piping that incorporate the 60Co are dissolved with reductive and oxidative chemical reagents. The SS base metal of the piping is exposed to reactor water after the chemical decontamination and the growth rate of the oxide film that incorporates the 60Co of the piping during plant operation just after the decontamination is higher than that just before it. Therefore, there is a possibility that the deposition amount of 60 Co on the piping just after decontamination is higher than that just before the chemical decontamination. Actually, rapid deposition amount increases of 60Co within a few operating cycles have been observed in some nuclear power plants. Then, we developed the Pt coating (Pt-C) to lower the recontamination by 60Co after the chemical decontamination. In the Pt-C process, a Pt layer is formed in an aqueous solution on the SS base metal of the piping using sodium hexahydroxyplatinate (IV) and hydrazine. In this study, we confirmed that the suppression effect by the Pt-C toward 60Co deposition on SS using a 60Co deposition test under hydrogen water chemistry. The deposition amounts of 60Co which were incorporated in oxides after 1000 h with and without Pt-C process were about 90 and 10.2 Bq/cm2, respectively. The amount of 60Co deposition with Pt-C is about 10% that of non-coated specimens. 16 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-115 Co-60 DEPOSITION ON CARBON-STEEL STRUCTURAL MATERIALS AFTER SEAWATER INFILTRATION IN BWR PLANT Hiromitsu Inagaki1, Osamu Shibasaki2#, Koji Negishi2, Yumi Yaita2, sato Okamura2, Yutaka Uruma2, Seiji Yamamoto2, Hajime Hirasawa3 1 Chubu Electric Power Co., Inc., 5561, Sakura, Omaezaki, Shizuoka, 437-1695, Japan 2 Toshiba Corporation, 4-1, Ukishima-cho, Kawasaki-ku, Kawasaki, 210-0862, Japan 3 Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523, Japan # Corresponding author: [email protected] Abstract Seawater infiltration occurred during shutdown of the Hamaoka Unit 5 (H-5). Chloride ion (Cl-) is known to affect the corrosion behavior of carbon steel, and it may change the properties of the oxide film formed on the surface. Co-60 deposition in high-temperature water is strongly related to the oxide film properties, and any change in the properties may affect the Co-60 deposition after the plant is restarted. This paper shows the results of Co-60 deposition tests of carbon steels under simulated H-5 water conditions. Specimens for the Co-60 deposition tests were prepared in three steps, which simulated the conditions of normal plant operation, seawater infiltration, and chemical decontamination after the infiltration. The first step was a prefilming step under Normal Water Condition (NWC). The second step included two different conditions: seawater infiltration and keeping after infiltration. Prefilmed specimens were immersed in 450 ppm Cl- diluted artificial seawater at 513 K for 24 hours. Following that, the specimens were immersed in 50 ppm Cl- diluted artificial seawater at 323 K for 100–500 hours. During the second step, the prefilming oxide (NiFe2O4) flaked off in spots. In the third step, the oxide remaining on some specimens after the second step was removed chemically. The three types of prepared specimens, that is, a prefilmed specimen, an exposed specimen, and an oxide-removed specimen, were used for the Co-60 deposition tests using 0.015 Bq/cm3 Co-60 solution for 500 or 1000 hours under NWC conditions. After the deposition tests, the Co-60 activity was measured with a Ge detector. From the results of the deposition test, at the spots where flaking occurred in the second step, only loose hematite was formed, and generation of a new protective film was not observed. The amount of Co-60 deposited on the exposed specimen was more than that on the prefilmed and oxide-removed specimens. The simulated infiltrating conditions inhibited the regeneration of a protective film and caused an increase in the amount of Co-60 deposited. Keywords: Co-60, RI deposition, Seawater infiltration, Carbon steel, Chloride ion 17 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-162 STUDY ON THE DECONTAMINATION OF 400 SERIES STAINLESS STEEL SURFACES AND DISSOLUTION OF CHROMIUM SUBSTITUTED NICKEL FERRITES V. S. Sathyaseelan, A. L. Rufus, P. Chandramohan, H. Subramanian and S. Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities, Kalpakkam, Tamilnadu – 603 102, INDIA # Corresponding email: [email protected] Abstract Full system decontamination of Primary Heat Transport system (PHT) of Pressurised Heavy Water Reactors (PHWRs) is carried out using weak organic acids at about 90 °C. During this low temperature process, decontamination factors (DFs) achieved on carbon steel (CS) surface was quite good as it was effective in dissolving magnetite formed on CS surfaces. However, the DF achieved on stainless steel (SS) and other non- CS surfaces were not that appreciable as the process was not effective in dissolving Cr and Ni substituted oxides present on these surfaces. “End fittings” surface was one such area where low DF was achieved. SS403 and SS-410 are the material of construction of “End Fittings” and “End Fitting Liners” respectively. Hence, to develop an effective decontamination process for “End Fittings” material surfaces, studies were carried out with SS-403 and SS-410 specimens. Passivated SS-403 and SS-410 surfaces were prepared by exposing the specimens under simulated PHWR PHT system chemistry conditions. The oxide film was characterised by chemical and physical techniques such as XRD, Raman spectroscopy and SEM-EDX. Three formulations evaluated for the dissolution of the oxide films formed over these alloys were i) Two-step process consisting of oxidation and reduction reactions, ii) Dilute Chemical Decontamination (DCD) and iii) high temperature Process at 160 °C. Material compatibility study also was carried out in these three formulations. The two-step and high temperature processes could dissolve the oxide film completely. But, the DCD process could remove only 60%. The twostep process is time consuming and generates large quantity of waste. Whereas, the high temperature process is less time consuming and generates only low volume of waste. Hence, high temperature process is recommended for SS decontamination. The high temperature process was evaluated for the dissolution of Cr substituted Ni ferrites also. This type of oxides is formed over SS-300 series alloys. Ni substituted Cr ferrites of composition NiCr(1-x)Fe(2-x)O4 (x = 0, 0.2, 0.4, 0.6, 0.8 and 1) were prepared by combustion route, characterised by XRD and Raman spectroscopy. Dissolution study of these oxide powders was carried out in NTA formulation at 160 °C. The formulation was very effective in dissolving this oxide and the rate of dissolution decreased with the increase in Cr substitution. 18 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-161 EFFECT OF HIGH CONCENTRATION GADOLINIUM NITRATE IN REACTOR MODERATOR SYSTEM Debasis Mal, Puspalata Rajesh, S. Rangarajan, S. Velmurugan#, Water & Steam Chemistry Division, BARCF, Kalpakkam, India # Corresponding Author: [email protected] Abstract Gadolinium is used as a neutron poison in nuclear reactors to control the reactivity because it has high thermal neutron absorption cross section (~49,000 b) and good solubility in water. Gadolinium nitrate is added with nitric acid to the moderator heavy water and the pH is maintained in the range of 5.0 to 5.5 to prevent gadolinium precipitation. Usually the concentration of gadolinium (Gd3+) used is ~15 ppm during the actuation of secondary shutdown system. In the moderator system of a proposed tube type boiling water nuclear reactor of Indian origin, a higher concentration (20-400ppm) of soluble neutron poison, Gd(NO3)3 was proposed to be used in the emergency safety shutdown system. Effect of this high concentration of gadolinium nitrate in the reactor moderator is evaluated from the angle of generation of molecular products viz. H2 and H2O2 due to radiolysis. H2 yield was found to increase linearly with absorbed dose (10 - 100kGy). With increasing Gd concentration there was increase in H2 yield but the increase was marginal in 100 to 400 ppm range. Both the initial yield and saturated concentrations of H2O2 (at higher doses) in normal and off - normal conditions were also estimated. It was observed that the head space provided above the liquid phase in irradiation zone has a substantial effect on the generation of H2. With decreasing head space, H2 generation increased and went through a maximum. Production of H2O2 was also observed to be decreased in case of fully filled samples as compared to the ~60% filled cases. Radiolysis of Gd(NO3)3 in high purity D2O was carried out to see the isotope effect and D2 formation was observed to be lowered than H2 for same Gd(NO3)3 concentration solutions in light water. The above results were discussed in detail in this paper. 19 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-173 METAL ION IMPRINTED POLYMERS FOR EFFECTIVE RADIOACTIVE WASTE SEGREGATION IN NUCLEAR INDUSTRY Anupkumar Bhaskarapillai WSCD, Chemistry Group, BARC, Kalpakkam – 603102, India # Corresponding author: [email protected] Abstract Routine decontamination campaigns of nuclear reactors are generally effective in removing various radionuclides such as cobalt, caesium, etc., and bring down the radiation field. However, during some of the decontamination campaigns, the radiation field at some surfaces were seen to have actually gone up. This was found to be due to lack of removal of antimony isotopes by the regular ion exchange resins used, which subsequently deposited over out of core surfaces leading to increased radiation field on those surfaces. Thus there exists a need for efficient antimony removal system. We have earlier reported the synthesis of nano titania impregnated - epichlorohydrin crosslinked chitosan beads, which were shown to be capable of complete sorptive removal of antimony from its aqueous solutions of concentration ranging from 150 ppb to 120 ppm. In this study, in order to understand the sorption mechanism and to fine tune the bead composition, the effect of crosslinker concentration used in the synthesis on the swelling and sorption properties of the beads was investigated in detail. The variation effected significant changes in physical parameters such as bead diameter, swelling ratio, equilibrium water content, and true wet density. Sorption capacity, unlike with regular resins, was found to increase with increase in crosslinker amount. The antimony sorption capacity of the crosslinked beads prepared by crosslinking 0.3 g uncrosslinked beads with 6.4 mmol epichlorohydrin (crosslinker) was 493 µmol/g. Noncrosslinked beads showed a capacity of 75 µmol/g, while the crosslinked beads made with the least amount of crosslinker (0.64 mmol per 0.3 g beads) showed a capacity of 133 µmol/g. These results indicate the possible involvement of the crosslinker in the sorption 20 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-136 CATION DISTRIBUTION IN FERRITES AND ITS EFFECTS ON THE CHEMICAL DISSOLUTION BEHAVIOUR P.Chandramohan#, M.P.Srinivasan and S.Velmurugan Radiation and deposit control studies section, Water and steam chemistry division, Chemistry group BARC, Kalpakkam Tamilnadu-603102 # Corresponding email: [email protected], [email protected] Abstract Ferrites are formed on the steel surfaces as a protective corrosion oxide film on the heat transport surfaces in the water cooled nuclear reactors. These oxides film acts as a host to many neutron activated corrosion products (ACPs) leading to man-rem problem during the service maintenance. Understanding of chemical dissolution kinetics of these ferrites is important aspect in the development of decontamination process with aim of good decontamination factors. Ferrite shows a cation distribution as a function of parameter like metal ion substitution, crystallite size and temperature. Change in the cation distribution in ferrite can effect its dissolution process. The following three ferrites namely CoFe2O4/ZnFe2O4/MgFe2O4 were studied for its chemical dissolution behaviour as a function of the cation distribution. CoFe2O4, MgFe2O4 and ZnFe2O4 shows an inversion parameters of 0.95, 0.46 and 0.06 respectively. The above ferrites with different cation distribution were achieved by the thermal treatment. The variation of cation distribution in ferrite was monitored/characterised by the Raman spectroscopy. Chemical dissolution of these ferrites were carried out in NAC formulation. Dissolution process was monitored by the metal ion dissolution in the solution. Dissolution data was fitted to the following two models ‘Shrinking sphere model’ and ‘Factual chain mechanism model’ to elucidate the kinetic parameter. We tried to establish correlation between the cation distribution in the ferrite and 80 the dissolution kinetics of ferrites. ZnFe2O4 (‘ ’= ~ 0.06) showed 𝑘𝑜𝑏𝑠(𝐹𝑒) = 1.250x10-3min-1 and 80 -3 -1 ZnFe2O4 (‘ ’= ~ 0.30) showed 𝑘𝑜𝑏𝑠 (𝐹𝑒) = 2.295x10 min , indicating ZnFe2O4 with high inversion parameter showed higher dissolution rate. Activation energy for the ZnFe2O4 (‘ ’= ~ 0.30) and ZnFe2O4 (‘ ’= ~ 0.06) in NAC formulation was 58.4 and 61.5 kJ mol-1 respectively. CoFe2O4 and MgFe2O4 also showed the dependence of its chemical dissolution behaviour on the cation distribution. From the above study we could establish effect of cation distribution on the dissolution kinetic of the ferrites 21 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-130 AREVA’S TOOLBOX FOR LONG-TERM BEST PERFORMANCE AND RELIABLE OPERATION OF NUCLEAR STEAM GENERATORS Andreas Drexler#, Steffen Weiss, Neil Caris, Christoph Stiepani# AREVA GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany) # Corresponding author: [email protected] Abstract Long-term integrity and high performance of major plant systems and components are of uppermost importance for the successful operation of any power plant. The objective of AREVA’s asset management program is to support operators by minimizing corrosion damage and performance losses of water-steam cycle systems and components and thereby to maximize the availability and economic performance of the plant. AREVA’s experience gathered with water-steam cycle chemistry treatments in more than 40 years yields the conclusion: accumulation of corrosion products in steam generators may result in local overheating and enrichment of impurities up to critical levels. This can lead to several degradation phenomena of the structural materials of the steam generators. Therefore, minimization of corrosion product generation and prevention of deposit accumulation is the one main goals of water chemistry. The AREVA approach for long term best performance and reliable plant operation consists of: • Control measures (e.g. water chemistry, SG cleanliness “fouling factor”, sampling system assessment, aging management ) • Corrective measures (e.g. chemical and/or mechanical cleaning, component replacement, coating) • Preventive measures (e.g. pH optimization, innovative additives like film forming amines, technical support for optimized plant operation ) Such asset management program is in principle a closed cycle process. A detailed technical assessment of the current situation is only a first step. In the subsequent steps appropriate measures which improve the current status or counteract on identified issues are identified and applied. These corrective and/or preventive measures cover a wide range like improvements of water chemistry treatments, primary and secondary side mechanical and/or chemical cleanings and changes of the material concept. This paper describes AREVA’s approach and the according field experiences of this asset management program. 22 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-140 TWO DECADES OF EXPERIENCE WITH STEAM-WATER CHEMISTRY MAINTENANCE OF FAST BREEDER TEST REACTOR K.Ganapathysubramanian, A.Suriyanarayanan≠ and B.S.Panigrahi Reactor Chemistry Section, Reactor Operation and Maintenance Group Indira Gandhi Centre for Atomic Research, Kalpakkam #Corresponding author: [email protected] Abstract Fast Breeder Test Reactor (FBTR) at Kalpakkam is a 40 MWt, loop type, sodium cooled fast reactor. The fission heat generated in the core is extracted by primary sodium circuit and the thermal energy is transferred to non-radioactive liquid sodium in the secondary circuit which in turn, heats Once Through-type shell and tube counter current Steam Generator (OTSG) for producing super heated steam at 480C and 125 kg/cm2. This secondary circuit is provided to avoid the ingress of hydrogenous materials and pressure surges reaching the core in the event of SG tube leak. Corrosion related problems are very less in the sodium circuits due to the absence of electrochemical reaction. The OTSG consists of four modules each of 12.5 MWt rating. OTSG was chosen due to its higher thermal efficiency and lesser inventory of steam/water in OTSG as it reduces the severity of sodium-water reaction, in case of tube leak. From the point of view of corrosion and deposition, the chemistry specifications are more stringent for OTSG than those of drum type boilers because 100 % conversion of feed water into steam takes place in OTSG. The chemistry requirements are achieved by providing ion exchange resin based online condensate polishing to remove ionic and suspended impurities. Dissolved Oxygen and pH are maintained by all volatile treatment (AVT) using hydrazine and ammonia respectively. Being a test reactor, a dump condenser with 100 % steam dump facility with cupro-nickel tubes is available for uninterrupted reactor operation during the non-availability of turbine. Regenerative feed heating by the exhausted steam from the turbine is also available to stage heaters and deaerator. Efficient water chemistry control plays important role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. This paper describes the operational difficulties such as premature exhaustion of CPU, impurity pick up from the system, silica excursion, online monitoring and suitable modifications carried out in the circuit for improvement. 23 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-105 IDENTIFICATION OF BOILER TUBE LEAK IN RAPS-2 BY MEASURING IODINE-134 ACTIVITY IN BOILER WATER SAMPLE OF RAPS USING OF GAMMA SPECTROMETRIC TECHNIQUES. P. K. Pal# and R. C. Bohra Rajasthan Atomic power station -1&2, Kota (Rajasthan) # Corresponding author: [email protected] (Fax 01475-242274) Abstract The boiler tube made up of Monel-400 of RAPS-2 has failed on few occasions. Due to the failure of boiler tube the active heavy water enters into boiler and feed water leading to contamination of radioactivity in secondary water circuit. The identification of boiler tube failure was done by measuring activity of Iodine-134 in the boiler water sample using gamma spectrometry using high purity germanium detector. In order t to increase the sensitivity of the method 5 liters of Boiler water sample was passed through a plastic column containing 40 ml of anion resin & 10 ml of activated charcoal to capture the isotopes of Iodine in the anion resin. Samples were collected from all 8 Boilers of RAPS-2. The activity of I-134 was shown only by Boiler #5. No other boilers showed any activity of I-134. This indicated that Boiler #5 had leaky tubes. The leaky hairpin of boiler #5 was identified by measuring Tritium, IP & I-131 activity in the riser and down comer of all 10 HXs. On the basis of Trium and IP result, HX-7 was identified as leaky hairpin. 24 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-170 IMPROVE STEAM GENERATOR MOISTURE CARRYOVER RATE AT MAANSHAN NPS BY CLEANING STEAM DRUM INTERNAL SLUDGE Meng-Jen Chen Taiwan Power Company Maanshan Nuclear Power Station #387 nanwan Rd. Hengchung Pingtung Taiwan Phone: +886 8 8893470~2810 Fax: +886 8 8894817 # Corresponding author: [email protected] Abstract 2013 August Maanshan Nuclear Power Plant commissioned perform steam generator moisture carryover test (MCO) and get a high rate of both unit. The reported MCO values for the unit 2 SGs significantly higher and thus more urgent to adress , as the average MCO value of 0.31% is substantially higher than the design limit and what is considered acceptable(0.25%) by most turbine vendors. With both unit MCO beyond the design limit, a plan needs to be developed to determine the cause of these high values(via an inspection of the steam drum region of the SG , and then develop actions necessary to improved the MCO rate). Westinghouse made steam drum inside inspection and report, there is no obvious regional water separator device degradation phenomena, resulting in MCO phenomenon is due to sludge accumulation in the dryer equipment bend, causing the dryer device function is reduced. Maanshan nuclear plant decided using chemical and join some manhand process to remove sludge in the dryer at NOV-2013 (unit 1EOC-21). Washing steps are as follows:1. manually remove visible mud inside the steam drum.2. Loose soaked with chemicals and solvents to wash portion of sludge.3. Low pressure water flush with the bottom of the dryer to remove loose sludge Results: Step 1 manually remove visible mud (SG A: 14kg, B: 16.5kg, C: 19.8kg) Step2 Chemical dissolve and remove sludge from the steam generator (SG A: 149.8kg, B: 213.9, C: 248.9kg) Step 3 flush and collect insoluble sludge (SG A: 97.3kg, B: 93 kg, C: 50.1kg) After steam drum washing process, Maanshan NPS check the photo from micro cameras and find most of the sludge was remove from the dryer vane pocket, but we still need to perform MCO test to confirm the result. 25 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-147 ENVIRONMENTAL IMPACT OF CONDENSER EFFLUENTS INTO COASTAL MARINE ENVIRONMENTS: NEED FOR CONTINUOUS MONITORING V. P. Venugopal Biofouling and Biofilm Processes Section, Water and Steam Chemistry Division, Bhabha Atomic Research Centre, Kalpakkam, India 603102. # Corresponding Author: [email protected] Abstract: Electric plants working on the principle of steam-water cycle require large amounts of water for condenser cooling purpose. Nuclear power plants require, on an average, about 3 m3 cooling water per minute per megawatt of electricity generated. Owning to the scarcity of large sources of freshwater for cooling, newer power plants, particularly in water-stressed parts of the world, tend to get located in coastal regions, where they can make use of the abundant seawater. However, this also poses a problem, in terms of the biofouling potential of coastal marine environments. Sessile benthic organism, which are generally present as part of the coastal marine ecosystem, extend their habitat into the cooling water system of the power plant. It is often observed that massive growth of such fouling organisms may endanger normal operation of the cooling water system, unless appropriate control measures are adopted. Presence of calcareous organisms such as mussels and barnacles in the precondenser sections of the power plant is a common sight; but these organisms, when lodged inside condenser tubes, can not only reduce the heat transfer efficiency but also can cause localized corrosion and tube leakage, leading of ingress of seawater into the steam-water system. It is, therefore, important that appropriate control measures are adopted to discourage the growth of the organisms. However, this needs to be done in an environmentally sustainable manner, as the cooling water is ultimately discharged back into the sea. The presentation aims to give and overview of the biofouling problems generally encountered in a typical tropical coastal power station operating in India and the chemical control measures adopted and their effectiveness. The talk also throws light on the more recent advances in biofouling control such as surface modification and use of nanotechnology which, in the foreseeable future, may provide more lasting and environmentally sustainable solutions. 26 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-149 NANOPHASE MODIFIED FLY ASH CONCRETE WITH SUPERIOR CONCRETE PROPERTIES, DURABILITY AND BIOFOULING RESISTANCE FOR SEAWATER APPLICATIONS Vinita Vishwakarmaa#, R.P. Georgeb U. Sudhaa, D. Ramachandrana, Kalpana Kumaric, R. Preethac, U.Kamachi Mudali b and C. S. Pillaic a Centre for Nanoscience and Nanotechnology, Sathyabama University, Chennai-600119 b Corrosion Science and Technology Group, IGCAR, Kalpakkam-603102 c Civil Engineering Group, IGCAR, Kalpakkam-603102 # Corresponding Author: [email protected] Abstract There are many concrete structures in the cooling water system of nuclear power plants that are exposed to seawater in the form of tanks, pillars and reservoirs. These structures come in contact with aggressive chlorides and acid producing microbes and deteriorate by chemical and biological factors. Recently fly ash (FA) concrete has emerged exhibiting excellent degradation resistance in seawater environments. However some disadvantages are reported like lesser early strength, higher carbonation and calcium leaching. This work attempted to modify FA concrete by adding nanoparticles of TiO 2 and CaCO3 for increased strength and degradation resistance. Four types of concrete and mortar mix namely fly ash concrete (FA), FA with 2% TiO2 nanoparticles (FAT), FA with 2% CaCO3 nanoparticles and FA with 2% TiO2: CaCO3 nanoparticles were cast and immersed in seawater for a year. Strength and durability were evaluated using parameters like compressive strength, split tensile test, Rapid chloride permeability test (RCPT), half cell potential test (HCP),carbonation test and pH degradation. Detailed biofilm characterizations were attempted using microbiological and molecular biology tools to study the antibacterial properties. Calcium leaching and sulfate attack studies were carried out by laboratory exposure studies. Using field emission scanning electron microcopy, EDAX and X-ray diffraction technique (XRD), microstructural properties and chemical phases were identified. All the nanophase modified FA specimens showed superior properties compared to FA concrete with respect to strength, carbonation depth, calcium leaching and antibacterial activity. Results are discussed in detail in the paper. 27 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-151 CONTROL OF BIOFOULING ON TITANIUM CONDENSER TUBES WITH THE USE OF ELECTROLESS COPPER PLATING B. Anandkumar1, R.P. George1*, D. Ramachandran2, U. Kamachi Mudali1 1 Corrosion Science and Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam – 603102, India 2 Centre for Nanoscience and Nanotechnology, Sathyabama University, Chennai – 600119, India *Corresponding author: [email protected] Abstract In sea water environments titanium condenser tubes face serious issues of biofouling and biomineralization. Electroless plating of nanocopper film is attempted inside the tubes for the control of biofilm formation. Using advanced techniques like AFM, SEM, and XPS, electroless copper plated flat Ti specimens were characterized. Examination of Cu coated Ti surfaces using AFM and SEM showed more reduction in the microroughness compared to anodized Ti surface. Cu 2p3/2 peak in XPS spectral analysis showed the shift in binding energy inferring the reduction of the hydroxide to metallic copper. Tubular specimens were exposed to sea water up to three months and withdrawn at monthly intervals to evaluate antibacterial activity and long term stability of the coating. Total viable counts and epifluorescence microscopy analyses showed two orders decrease in bacterial counts on copper coated Ti specimens when compared to as polished control Ti specimens. Molecular biology techniques like DGGE and protein expression analysis system were done to get insight into the community diversity and copper tolerance of microorganisms. DGGE gel bands clearly showed the difference in the bacterial diversity inferring from the 16S rRNA gene fragments (V3 regions). Protein analysis showed distinct protein spots appearing in electroless copper coated Ti biofilm protein samples in addition to protein spots common to both the biofilms of Cu coated and as polished Ti. The results indicated copper accumulating proteins in copper resistant bacterial species of biofilm. Reduced microroughness of the surface and toxic copper ions resulted in good biofouling control even after three months exposure to sea water. 28 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-155 BIOFOULING COMMUNITY PATTERN ON VARIOUS METALLIC SURFACES IN THE COASTAL WATERS OF KALPAKKAM, SOUTHWESTERN BAY OF BENGAL Gouri Sahu, K.K. Satpathy*, A.K.Mohanty and V.K.Bindu Environment & Safety Division, EIRSG, Indira Gandhi Centre for Atomic Research, Kalpakkam-603 102, Tamil Nadu, India *Corresponding author: [email protected] Abstract Biofouling causes great operational hazard in different marine installations across the globe. And the expenditure incurred on combating biofouling is astounding. It is reported that shut down of a 235 MW (e) power station due to fouling, costs about 170 lakhs (at Rs. 3.00 per kw/h) per day. Because of this economic implication, biofouling has been a thrust area of study for the marine researchers. To assess the biofouling pattern, metallic surfaces are the best options because of their extensive use at various installations in the marine environment. Hence, knowledge on qualitative and quantitative aspects of biofouling with respect to metal surfaces is of great value to design an efficient fouling control strategy. Keeping this in mind, nine types of metal [SS-316, SS-304, MS, Titanium, Admiralty Brass, Aluminum Brass, Copper, Monel and Cupro-nickel] panels (12 x 9 x 0.1 cm) were exposed to coastal water of Kalpakkam from MAPS jetty at a depth of 2 m below the lowest low tide. Results indicated that copper based panels were found to be foul-free except monel. Although, fouling settlement was encountered on monel, the adherence was weak. Non-copper based metals showed 100% area coverage with high population density. However, in case of MS, due to exfoliation of corrosion deposits, unevenness in fouling colonization at later stages of development took place, though the early settlement was unaffected by initial corrosion. As expected, Titanium showed high rate of fouling growth along with high fouling diversity compared to other noncopper based metals. Absence of specific foulants such as, crustaceans and algae on Titanium surface reported by others was not observed during our study. The information on Titanium would be handy for Prototype Fast Breeder Reactor (PFBR) cooling water system wherein, the same has been selected as condenser and process water heat exchanger material. For non-copper based alloys including monel the fouling load ranged from 18 to 40 g. 100 cm-2. The major fouling organisms such as, barnacle, green mussel and ascidian constituted ~ 70-80% of the total fouling. In the present study, sequence of fouling succession was as follows, barnacle – hydroid - sea anemone – ascidian and finally green mussel (Perna viridis Linn. 1758). The paper also discusses species diversity indices (diversity, richness and evenness) in detail. 29 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-109 PLASMA NITROCARBURIZING PROCESS –A SOLUTION TO IMPROVE WEAR AND CORROSION RESISTANCE Alphonsa Joseph#, Ghanshyam Jhala, and S. Mukherjee FCIPT, Institute for Plasma Research, Gandhinagar, Gujarat, INDIA. # Corresponding author: [email protected], [email protected] Abstract To prevent wear and corrosion problems in steam turbines, coatings have proved to have an advantage of isolating the component substrate from the corrosive environment with minimal changes in turbine material and design. Diffusion based coatings like plasma nitriding and plasma nitro carburising have been used for improving the wear and corrosion resistance of components undergoing wear during their operation. In this study plasma nitrocarburising process was carried out on ferritic alloys like ASTM A182 Grade F22 and ATM A105 alloy steels and austenitic stainless steels like AISI 304 and AISI 316 which are used to make trim parts of control valves used for high pressure and high temperature steam lines to enhance their wear and corrosion resistance properties as shown in Fig. 1. The corrosion rate was measured by a potentiodynamic set up and salt spray unit in two different environments viz., tap water and 5% NaCl solutions. The Tafel plot of ASTM A182 grade F22 steel shows that plasma nitrocarburising for 6 and 24 hours show better corrosion resistance compared to that of the untreated steel ( Fig. 2). Fig. 1: Control valve components mounted Fig. 2: Tafel plot of untreated (UT) for plasma nitrocarburizing. plasma nitrocarburized F22 steel for 6 (PNC6) and 24 (PNC24) hours It was found that after plasma nitrocarburizing process the hardness of the alloy steels increased by a factor of two. The corrosion resistance of all the steels mentioned above improved in comparison to the untreated steels. This improvement can be attributed to the nitrogen and carbon incorporation in the surface of the material. This process can be also applied to components used in nuclear industries to cater to the wear and corrosion problems. 30 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-154 FEASIBILITY STUDY OF A NON-CHEMICAL TECHNIQUE FOR FOULING CONTROL Yaw-Ming Chen Industrial Technology Research Institute, Material and Chemical Research Laboratories, # Corresponding Author: [email protected] Abstract In nuclear power plants, fouling occurred in different systems and caused operation problems. Many factors affect the behavior of fouling. Among them, zeta potential of particles suspended in a liquid plays an important role in deposition of particles onto surface. In our work, a non-chemical water treatment based on the effect of zeta potential was tested to verify its effectiveness to reduce fouling. 31 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-121 ENVIRONMENTAL SUSTAINABILITY BY ADOPTION OF ALTERNATE COOLING MEDIA FOR CONDENSER COOLING Jaymin Gandhi, Nilesh Patel Adani Infra India Ltd. Ahmedabad,Gujarat, India. # Corresponding author: [email protected], [email protected] Abstract Water having ability to dissolve most substances and to support biological life, every cooling water system in power plant is subjected to potential operational problems which are mainly corrosion, scaling and biological fouling. Control of cooling water chemistry is very critical in preventing above said problems. In view of scarcity of water and looking into the future trends in the environment protection, water media can be replaced with air. Having such concept in thermal & combined cycle power plants, use of Air-cooled condenser(ACC) for Nuclear power plant may be explored. During last decade number of installations with ACC also increased, largely in response to the growing attention being paid to environmental concerns as well of water scarcity. The rising importance of ‘Save Water & Environment’, calls for a broader understanding of the design and application principles involved for ACC.This paper identifies the basic configurations of air cooled condensers used in the power industry together with their merits & demerits when compared to those exhibited by traditional steam surface condensers including environmental and corrosion issues. Several factors that affect the performance of air-cooled condensers are described in detail, especially the consequences that result from the fouling of the finned-tubes. To rectify the degradations in performance that result from external tube fouling, a number of cleaning procedures are described. Due to relatively high cost of sweet water and large requirement of sea water, Air cooled condenser may become viable option in future. Keywords: Air-cooled Condensers, Environmental aspect 32 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-103 RADIOLYSIS OF WATER AT HIGH TEMPERATURE AND PRESSURE CONDITIONS: A PICOSECOND PULSE RADIOLYSIS EXPERIMENT AND NUMERICAL SIMULATIONS Yusa Muroya1*, Tetsuro Yoshida1, Yosuke Katsumura2,3, Shinichi Yamashita3, Mingzhang Lin4, Takahiro Kozawa1 1 Institute of Scientific and Industrial Research, Osaka University, Mihogaoka 8-1, Ibaraki, Osaka 567-0047, Japan 2 Japan Radioisotope Association, Honkomagome 2-28-45, Bunkyo, Tokyo 113-8941, Japan 3 School of Engineering, University of Tokyo, Hongo 7-3-1, Bunkyo, Tokyo 113-8656, Japan 4 School of Nuclear Science and Technology, University of Science and Technology of China, 96 JinZhai Road, Hefei, Anhui 230026, P.R. China *Corresponding author: [email protected] Abstract Radiolytic products of coolant material (light water) under strong radiation field in RPV are known to give undesirable effects on nuclear structural materials. Understanding of the fundamental processes will be of great importance on powerful support to various application fields in water chemistry. Ionization and excitation of water molecules by ionizing radiations initiate very fast physical and chemical processes within μs(10-6 s), ns (10-9 s) or even ps (1012 s), prior to formation of primary radiolytic species (e-aq, OH, H, H2, H2O2 etc.). Through the processes, the radiation chemical yields (G-values) are supposed to change dynamically depending on time and also on temperature. However, because of so high reactivity (short lifetime), it was difficult to observe experimentally the temporal behaviors (spatially inhomogeneous reactions, which is called spur diffusion reactions). In this work, the fundamental processes (G-values of the intermediates and the fast reaction kinetics) of the radiolysis of water at high temperature and pressure conditions were investigated by a newly developed ultrafast (picosecond time-resolved) pulse radiolysis system, and also by numerical analyses such as the Monte-Carlo simulation and the Spur diffusion kinetic model simulation. 33 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-104 INJECTION OF NANO-PARTICLES IN MITIGATING FLOW ACCELERATED CORROSION (FAC) DAMAGE IN THE SECONDARY SYSTEM OF NUCLEAR POWER PLANTS (NPPS) Dong Seok Lim#, Hee Kwon Ku, Jae Seon Cho FNC Tech., Heungdeok IT Valley, Heungdeok 1-ro, Giheung-gu, Yongin-si, Gyeonggi-do, 446-908, S. Korea # Corresponding author: [email protected] Abstract NPPs produces electric energy through phase transition of water. According to this, a piping, which is flow path, integrity is essential for safety functions. Erosion, FAC and fittings are corrosion failure mechanism by increasing service life. Especially, there are 10-kilometers of piping in secondary systems. It needs to estimate FAC and apply periodic management. Iron oxides produced by FAC cause power reduction and Loss Of Coolant Accident (LOCA) will be occurred through the continued piping wall thinning. In this study, corrosion rate of pipe materials with carbon steel(SA106.Gr.B) and low-alloy steel(SA335.P22) was evaluated for pipe configuration and dissolved oxygen concentration on 150℃, pH 9.5~10.0 and flow velocity of 5m/s. Temperature of 150℃ is well known that causes high FAC rate and pH consider a NPPs in-service condition. Further corrosion rate test was performed to develop FAC reduction technology through Pt-nanoparticle injection.In this study, corrosion rate is evaluated by weight depletion method. The results of material impact assessment show that corrosion rate of carbon steel is more higher than that of low-alloy steel because of Cr content. And also, the results of pipe configuration test show that case with 90° elbow had maximum wall thinning than with 180° horizontal pipe. The dissolved oxygen concentration test shows that low oxygen condition, ≤5ppb, had high corrosion rate compared to normal condition and the corrosion rate decreased 50% at Pt-nanoparticle injection test on maximum corrosion rate condition compared to maximum wall thinning condition without Ptnanoparticle injection. In this study, samples provided by each test case had analyzed through SEM-EDS(Scanning Electron Microscopy-Energy Dispersive X-ray Spectroscopy) and XRD(X-ray diffractometer). Behavior evaluation for oxide film was performed and Electrochemical corrosion potential(ECP) was measured for electrochemistry evaluation. To apply Pt-nanoparticle injection technology on nuclear power plant, study on injection conditions and methods are on-going. Keyword: nuclear power plant, corrosion, flow accelerated corrosion (FAC), pourbaixdiagram, carbon steel 34 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-106 ADDITION OF OXYGEN IN THE INLET OF RECOMBINER UNIT IN MODERATOR COVER GAS SYSTEM TO FACILITATE RECOMBINATION OF DEUTERIUM AND OXYGEN TO BRING DEUTERIUM CONCENTRATION IN SAFE LIMITS P. K. Pal# and S. Mukherjee2 Rajasthan Atomic Power Station-1&2 PO: Anushakti, Via: Kota, Rajasthan, 323302 2 STE (N), RR Site, Unit-1&2, Kota, Rajasthan # Corresponding author: [email protected] Abstract In moderator system of a PHWR, radiolytic decomposition of Heavy Water take place in the Calandria and D2 and O2 are formed. Since the mixture of D2 & O2 is explosive, there is a level and various action levels for concentration of Oxygen and deuterium in moderator cover gas. The maximum percentage limits of deuterium are 4% v/v in presence of Oxygen present in stoichiometric ratio. In March -2013, the deuterium concentration in moderator cover gas of RAPS-2 was increased to 3.0 % v/v and the oxygen concentration was only 0.93% v/v which was much less than the stoichiometric value (1.5%) . So it was decided to add oxygen in the inlet of recombiner unit of moderator cover gas system. As in year 1999 there was fire incident in Dalinton unit #3 due to combustion of EDPM which was used as seat material in the isolating valve in the inlet of oxygen pressure regulating valve. EDPM has low ignition temp. So Oxygen addition was not practiced in any other reactors. Oxygen addition was carried out in RAPS-2 with all precaution and with proper planning. Initially oxygen was added with very slow rate of 1 SCFH (Standard Cubic Feet per Hour) intermittently &the process was repeated to see any harmful effect. After qualifying the procedure, oxygen addition was done for 20 hrs at the rate of 2.5 SCFH and D2 concentration came down to 1.95 % v/v. This paper will consists of radiolytic decomposition of D 2O. The qualification plan for O2 addition, the data of moderator cover gas and moderator system parameters before, during and after Oxygen addition. 35 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-107 DETERMINATION OF MOISTURE CONTENT IN STEAMS BY ANALYZING SODIUM CONTENT IN STEAM GENERATOR WATER & STEAMS CONDENSATE OF A NUCLEAR POWER PLANT USING ION CHROMATOGRAPHIC TECHNIQUE AT DIFFERENT LEVELS OF BOILER WATER P.K.Pal# and R.C.Bohra Rajasthan Atomic Power Station-1&2 PO: Anushakti, Via: Kota, Rajasthan PIN code: 323302 # Corresponding author: [email protected] Abstract Dry steam with moisture content less than <1% is the stringent requirements in a steam generator for good health of the turbine. In order to confirm the same, determination of sodium is done in steam generator water and steam condensate using Ion Chromatographic techniques. Depending on the carryover of sodium in steam along with the water droplet (moisture), the moisture content in steam was calculated and was found to be < 1%, which is the requirements of the system. The paper described the salient feature of a PHWR, principle of Ion chromatography, chemistry parameters of Steam Generator and calculation of Moisture content in steam on the basis of sodium analysis at different boiler levels. The moisture content in steam increases with Boiler levels. So for smooth operation of Turbine an optimum level of 50 cm was selected for RAPS-2. In the abstract of the paper is accepted, please acknowledge the same such that the actual paper can be sent for publication in time. Please give the e mail id for easy and quick communication. 36 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-108 EXPERIENCE ON KKNPP VVER 1000 MWE WATER CHEMISTRY S. Ganesh, S. Selvaraj, M.R. Balasubramanian, P. Selvavinayagam#, Suresh Kumar Pillai, Kudankulam Nuclear Power Project, Tirunelveli Dist, Tamil Nadu- 627120 # Corresponding author: [email protected] ABSTRACT: Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration w*ith Russian Federation at Kudankulam In Tirunelveli District, Tamilnadu. Unit #1 attained criticality on July 13th 2013 and the unit was synchronized to grid on 22nd October 2013.This paper highlights experience gained on water chemistry regime for primary and secondary circuit. Primary Circuit: Primary circuit is a weekly Alkaline, reducing, Ammonia and Potassium water chemistry coordinated with Boric Acid. The structural material of reactor pressure vessel is low alloy steel 15X2HMФA (C‐ 0.15%, Cr‐2%,Ni‐1%, Mo‐ <1.0%, V‐<1.0%) cladded with austenitic stainless steel 08X18H10T (C‐0.08%, Cr‐18%, Ni‐10.0%, Ti‐ <1.0%). The other pipelines are low alloy steel cladded with austenitic stainless steel. One of the special features in the primary circuit of KKNPP is that four high temperature titanium filters are connected across reactor coolant pump to remove undissolved activated corrosion impurities. KOH is added as an alkalizing agent instead of LiOH to maintain pH in the primary circuit. This is being carried out to prevent tritium build up in the primary circuit. Ammonia is being added to maintain the dissolved oxygen and dissolved hydrogen to maintain reducing condition in the primary circuit. The control parameters and diagnostic parameters stipulated in the design are to ensure the design service life of the primary system equipments, low radiation build up on the out of core surfaces, minimum deposition and oxidation on the fuel clad. The experience on water chemistry of primary circuit is elaborated in this paper. Secondary Circuit: The structural material of SG body and collectors is of low alloy steel (perilitic class steel) 10GH2MΦA (C‐0.1%, Ni‐2%, Mo‐<1.0%, V‐<1.0%). Collectors are cladded internally with corrosion resistant austenitic stainless steel. Steam Generator, Low and high Pressure heater tubes are made of austenitic stainless steel, 08X18H10T (C‐0.08%, Cr‐18%,Ni‐ 10.0%,Ti‐ <1.0%). Steam Generators are horizontal which is unique in KKNPP provide saturated steam with very low moisture carry over. Online Cationic conductivity measurement to find out trace level impurity ingress into condensate system is also an unique feature in KKNPP. Secondary circuit pH and low dissolved oxygen is maintained by addition of ammonia and hydrazine hydrate respectively. The main objective of secondary water chemistry is to minimize the deposition and corrosion of the SG tube which is 3rd barrier in the defense in depth. Full flow condensate polishing unit and continuous SG blow down purification systems ensures the secondary water chemistry. The experience on water chemistry of secondary circuit is also elaborated in this paper. 37 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-110 KINETICS OF DISSOLUTION OF NI-CR CONTAINING IRON OXIDES SERIES (NICRXFE2-XO4) IN HMNO4 MEDIUM V. Balaji a, P. Chandramohan a, Ashish Tiwari b S. Rangarajan a, S. Velmurugana# a Water And Steam Chemistry Division, BARC Facilities, Kalpakkam - 603102, India b Chemistry Division, Trombay - 400085, India # Corresponding author: [email protected] Abstract The oxide film formed on steels in water–cooled nuclear reactors, has a duplex layer structure with an inner and outer layer of different composition and different microstructure. The films formed under Hydrogen Water Chemistry (HWC) are considerably thinner than those formed under Normal Water Chemistry (NWC). Ni-Cr ferrite is one of the important corrosion products formed on the structural material of steels in water–cooled nuclear reactors. The increase of chromium in the inner layers enhances the stability of the oxide. The dissolution of mixed oxides is considerable importance in terms of reducing radiation fields. In order to understand the mechanistic aspects of decontamination, a series of chromium substituted nickel ferrites with chemical composition NiCrxFe2.0-xO4 (0 x 2.0 in steps of 0.2) were prepared through sol-gel combustion route using different amounts of metal nitrates and urea/citric acid as the starting materials. The oxide precursors were annealed at 773 ± 5K for 4 h. The XRD and Raman patterns indicated that the synthesized oxides have single-phase spinel structure with crystallites in nano-size range. The oxidative dissolution of Ni-Cr ferrites were carried out in 3.0 mM HMnO4 medium at 363 ± 5K as a function of chromium substitution in the lattice. The rate constants were determined using both inverse cubic rate law and a general kinetic rate law models. The GKE model yields higher factor of increase, compared to ICR model. A similar trend of increasing rate constants as estimated by Ni /Cr release for the Cr-Ni ferrite (0.2 x 2.0) by both ICR and GKE models were also observed. The kinetic data on Ni-Cr ferrites dissolution in solution of HMnO4 were also analysed in terms of chain mechanism model which takes into account the changes in the fractional dimension of the surface of dissolving particles. 38 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-111 MEASUREMENT OF HENRY’S LAW CONSTANT IN HYDROGENATED LIOH/H3BO3 SOLUTION E. H. Lee, G. G. Lee, D. H. Lee, and D. H. Hur Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353, Republic of Korea Corresponding author: [email protected] Abstract In pressurized water reactors, hydrogen is added to the reactor coolant system in order to reduce the oxidation of water by radiolysis and to maintain reducing conditions. The dissolved hydrogen concentration in pressurized water reactors has been controlled within the range of 25~50cc (STP)/kg-H2O. It is well known that the dissolved hydrogen leads to primary water stress corrosion cracking. Therefore, the optimization of the hydrogen concentration in the reactor coolant system is regarded as one of several effective approaches to manage the material integrity and reduction of the radiation sources in the primary circuit. In order to predict the content of the dissolved hydrogen, it is needed to measure and monitor the content of the dissolved hydrogen accurately. In this work, the Henry’s law constant was experimentally determined in hydrogenated LiOH/H3BO3 solution. The Henry’s law constant was calculated from in-situ mesaurements of the hydrogen fugacity on the inside of the Pd-Ag tube for temperatures between 290℃ and 330℃ at pressures between 2000 psia and 2900 psia. The Henry’s law constant were found to decrease linearly with increasing temperature and with decreasing pressure. The observed trend in the Henry’s law constant corresponded well with other workers’, however, the absolute values were different from literature data. This may be due to the effect of vessel pressure on the hydrogen fugacity and the accuracy of the measuring sensor. To better calculate the Henry’s law constant, the pressure and the content of the dissolved hydrogen should be considered. In this work, the effects of temperature and pressure on the Henry’s law constant are described by an empirical model based on the experimental data. This emprical model is compared with other literature data. 39 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-116 SEASONAL VARIATION IN TRIHALOMETHANE LEVELS AT KALPAKKAM AND IN RELATION TO ORGANIC CARBON PRECURSORS R. Rajamohan1, V. P. Venugopalan#1 and Usha Natesan2 1 Biofouling and Biofilm Processes Section, Water and Steam Chemistry Division Bhabha Atomic Research Centre, Kalpakkam, Tamil Nadu 603 102, India 2 Centre for Water Resources, Anna University, Chennai, Tamil Nadu 600 025, India # Corresponding author: [email protected] Abstract Biofouling control in coastal power stations is generally achieved through the use of chlorination. However, chlorine reacts with natural organic matter, leading to the formation of several chlorinated by-products such as trihalomethanes (THMs). Environmental discharge of THMs is of concern, as these compounds have been reported to be carcinogens and mutagens. Madras Atomic Power Station (MAPS) employs continuous gas chlorination for controlling biofouling. Several studies have shown that the formation of THMs depends on several parameters, including the concentration of chlorine, total organic carbon (TOC), bromide, water temperature and pH. This study discusses the seasonal variation of TOC and compares it with the levels of chlorine residual in the formation of THM in the cooling water circuits of MAPS. The study showed that the more THM concentration was formed during the period when more concentration of chlorine was dosed and the formation of THM showed more correlation with the levels of chlorine than total orgainc carbon. Keywords: Chlorination by-products, Trihalomethanes, Total organic carbon, Total residual chlorine 40 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-117 ROLE OF REDUCTANTS IN CORROSION CONTROL OF MATERIALS RELEVANT TO NUCLEAR REACTORS Padma S.Kumar1, Sinu Chandran1, Puspalata Rajesh1, D.Mohan2, S.Rangarajan1 and S.Velmurugan#1 1 Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu, India 2 Membrane Lab, AC Tech., Anna university, Chennai, Tamilnadu, India # Corresponding author: [email protected] Abstract Presence of oxidants aggrevate corrosion of materials in aqueous medium. Removal of dissolved oxygen from aqueous systems of steam generating power plants, boilers etc. and inhibition of corrosion of component materials become very essential. Reductants like hydrogen gas or water soluble reducing agents can be used to control the corrosion and protect the structural components. Feasibility of using alternate reductants such as hydrazine, ammonium hydroxide and hydroxylamine which stays in liquid phase is studied in this paper. A comparative study of corrosion behavior of the materials Carbon steel, Stainless steel(SS304 LN), Monel-400 and Incoloy-800 in the oxidative and reductive conditions are being discussed. In nuclear industry, water radiolysis products like H2O2 is responsible for corrosion. Computation on the generation of oxidizing species (O2 and H2O2) and their distribution in steam and water phase were made. Analytical methods have been standardized to study the distribution of hydrazine, ammonia, hydroxylamine and hydrogen peroxide. Extensive studies were carried out at 90oC, 150oC to study the compatibility of structural materials in the oxidative and reductive environments. Electrochemical evaluation at 90oC showed that these reductants were very efficient to control corrosion of materials. Studies in HTHP system at temperature range 200 – 280oC evaluated the thermal stability of reductant hydrazine and its effect on redox potential of SS-304 LN. There was potential change from -0.4V (versus SHE) to -0.67 V (versus SHE) on addition of 5 ppm hydrazine at 240oC. The decomposition rate of hydrazine was observed to follow first order decay. As these reductants can be used in nuclear reactors, the radiation stability also was studied. The distribution of the oxidant hydrogen peroxide and the reductant hydrazine was found to be comparative in water-steam phases. The surface morphology of the exposed structural materials at high temperature were characterized by surface techniques such as SEM, EDX, RAMAN spectroscopy and by optical microscope. 41 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-119 IMPROVEMENT IN PERFORMANCE OF DM PLANT, SECONDARY SYSTEMS FOR ACHIEVING CHEMISTRY PERFORMANCE INDICATOR OF KGS-3&4 B.S.Sahu#, P.G.Raichur, M Srinivas and M P Hansora Kgs- 3&4, NPCIL, Uttar Kannada, Karnataka, India. # Corresponding author: [email protected] Abstract Kaiga Generating Station (KGS)-3&4 has two 220MWe Pressurized Heavy water Reactors. It uses Heavy water as moderator and coolant and DM (De-mineralized) water in secondary system for steam generation. Raw water for plant is taken from Kali River. Raw water is first treated in pretreatment plant and Dual media filter for turbidity removal. Chlorination is carried out for control of micro-organism. DM water is makeup to feed water which is the input to Steam Generator for production of steam for power generation. Continuous blow down through Boiler blow down (BBD) IX column is carried out to control Steam Generator (SG) chemistry. It was decided by NPCIL to calculate Chemistry Performance Indicator of KGS secondary and first time it was found 2.6 which was much higher than Standard and best achievable value of 1.0. Detailed analysis was carried out and improvements for DM plant, water treatment plant, BBD IX column, Steam Generator etc were identified. Turbidity of filter water was brought below 2.0 NTU. Many changes were incorporated in DM plant. Regenerate concentration, regeneration levels and regeneration procedures were modified. Resin replacement frequencies were fixed and brine treatment of anion resin was started at regular interval. For DM water production two mixed resin columns in series were used in place of earlier one mixed resin column. By these modifications DM water Chloride, Sodium and Sulphate were brought <1.0ppb from earlier 5-10ppb. Regeneration procedure of BBD IX column were standardized. Service life of BBD IX column was fixed and was isolated from service before complete exhaustion. Design deficiencies of BBD IX column was rectified by applying innovative idea. Online sodium analyzer war installed in boiler blow down line. By implementing these improvements Chemistry Performance Indicator of both units were brought down to 1.0, which is standard and best achievable value. 42 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-120 EXPERIENCE OF CHEMICAL TREATMENT FOR CONTROLLING CORROSION IN IDCT WATER OF KGS 3&4. V. Uday Kumar & B.S.Sahu# KGS- 3&4, NPCIL, Uttar Kannada, Karnataka, India. # Corresponding author: [email protected] Abstract KGS (Kaiga Generating Station)-3&4 is 220MWe pressurized heavy water reactor. Active Process Water Cooling system (APWC) cool active process cooling water through plate type heat exchanger. The heat from this system is dissipated to the atmosphere through Induced Draught Cooling Tower (IDCT). Continuous make up of system is carried out with raw water to compensate evaporation and blow down loss. Average Langlier Index (LI) of makeup water is -2.0. Cycle of concentration (COC) of APWC system water is around 4.0 and LI at this COC is around -0.1. As per design chlorination of water is carried out and 0.20.5ppm free residual chlorine (FRC) is maintained. Other than chlorination no chemical treatment was considered in the design. Considering that cooling water may have corrosion, scaling and bio-fouling problems, a detailed study was carried out. Corrosion, scaling and bio-fouling studies were carried out for three months by maintaining the COC around 4.0 & during this period normal chlorination was carried out. The results of the study had shown high corrosion rate for Carbon Steel (CS) but water did not have high scaling & the bio-fouling tendency. Sulphate Reducing Bacteria (SRB) and Total Bacteria Count (TBC) were evaluated & found within the limits. This indicated that water is corrosive in nature & a suitable chemical treatment needs to be carried out to control the corrosion of cooling water system. Chemical treatment in IDCT water with the formulation consisting of Zinc, Phosphonate, Azole and low molecular formulation was started along with chlorination. Biocide (Benzalkonium chloride) dosing was also started at regular intervals. After chemical dosing a downtrend trend of corrosion rate of CS was observed but still it was higher than limit. After increasing Zinc concentration in water from 0.2 to 0.5 ppm, CS corrosion was reduced to <2.0. 43 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-122 ELECTROCHEMICAL PASSIVATION STUDIES OF ZIRCALOY IN PRESENCE OF METAL ION Sinu Chandran, H. Subramanian, N. Sreevidya*, S. Rangarajan# and S. Velmurugan Water and Steam Chemistry Division, BARC Facilities, IGCAR Campus Kalpakkam, Tamilnadu, INDIA Telefax: +91 44 27480097 *MJS, MTD, Indira Gandhi Centre for Atomic Research Kalpakkam-603102 Tamilnadu, INDIA # Corresponding author:[email protected] Abstract Inorganic metal ion additives are being explored for controlling the corrosion and deposition of activated corrosion products on out of core surfaces in Pressurized Heavy Water Reactors. Addition of Mg2+ to the process stream is reported to be beneficial in reducing corrosion and corrosion product release from carbon steel. This added Mg 2+ ions can modify the oxides on other heat transfer surfaces and thus influence their corrosion behavior. In this context, an attempt was made to study the role of Mg2+ ions in modifying the passive films on Zircaloy surfaces by electrochemical passivation at ambient temperatures. Increased polarization resistance obtained from the impedance spectra recorded at OCP in the presence of Mg2+ ions revealed the beneficial effect of Magnesium. Different passive potentials were identified from the passive regions of the polarization curves. Each potential was applied to the specimen surface and the film growth was monitored by amperometry as well as by obtaining time evolution impedance spectra. The extent of Mg uptake in the oxide was evaluated by different surface characterization techniques. The results obtained from the above studies are discussed in detail in this paper. 44 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-123 OXIDATION/CORROSION COMPATIBILITY STUDIES OF P91 AND RAFM STEELS BY ELECTROCHEMICAL TECHNIQUES N Sreevidya, Sinu Chandran*, C.R. Das, S. K Albert#, S Rangarajan* and S Velmurugan* Material Technology Division, Indira Gandhi Centre for Atomic Research Kalpakkam-603102 Tamilnadu, INDIA *Water and Steam Chemistry Division, BARC Facilities, IGCAR Campus, Kalpakkam603102,Tamilnadu, INDIA # Corresponding author: [email protected] Abstract Oxidation behavior of Reduced Activation Ferritic Martensitic (RAFM) steel, the structural material for fusion reactors exposed to atmosphere for a period of ~6480 hrs has been studied giving special emphasis on the morphology and chemistry of the oxides formed. The results obtained are compared with those of Grade 91 (P91) steel from which the RAFM steel has been evolved mainly by replacing Mo and Nb with W and Ta respectively. RAFM corrodes faster than Grade 91 steel. Presence of chloride ions has a major role in deteriorating the corrosion resistance of RAFM steel. The oxides formed in both the materials consist of binary oxides of Fe and Cr, hematite, magnetite, chromia and Fe-Cr spinel oxides. Corrosion of these steels has also been studied by electrochemical techniques that confirm the inferior corrosion resistance of RAFM steel compared with that of Grade 91steel 45 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-124 PERFORMANCE RESTORATION TECHNIQUE DEVELOPED FOR FOULED HEAT EXCHANGER Dipankar Nanda1, Babloo Tiwari2, R. M. Pandey3#, Coolant Systems Lab (CSL) Raja Ramanna Centre for Advanced Technology, Indore, India # Corresponding author: [email protected] Abstract Heat exchanger (HE) is one of the important equipment of installations for satisfactory operation of power plants, chemical plants, Accelerator machine etc. The performance of HE depends on the material of construction (MOC) as well as good engineering practice adopted, and performance deterioration takes place due to surface deposition, making it a thermal insulator. In Indus-2 Electron Synchrotron Accelerator, RRCAT, the Plate Heat Exchanger (PHE) are installed to dissipate heat from primary process coolant (deionised water) to secondary cooling tower coolant (soft water) through parallel narrow passage of SS 316 corrugated HE plates. For achieving precise accerator beam stability, the process cooling water temperature stability is required to be maintained within ±10C. Deposition of scale takes place in secondary side of HE as Saturation Index (SI) is maintained at + 0.5. This affects the heat transfer coefficient. Hence, routine cleaning is required to remove the calcite scale of HE, leaving behind protecting layer of calcium carbonate scale on pipeline and other wetted parts of the loop to prevent corrosion. Unavoidable circumstances led to hard deposition of scales and the problem could not be even addressed by experts in this field. Samples were systematically analysed in the CSL laboratory to know the content of the deposit so that suitable method could be applied to selectively remove the foulants to finally clean the HE. About 48.52 % of deposit was found to be acid soluble, whereas approximately 44.14% of deposit dissolves in alkali. The remaining 7.43% residue was neither dissolved in acid nor in alkali which indicates that the undissolved part may be mostly of dust. The cleaning solution was formulated in-house to remove the scale from heat exchanger plates. Sulfamic acid solution at 800C was used to decompose calcium scale to liberate carbon dioxide, whereas sodium hydroxide solution with EDTA was used to remove remaining scale. The performance of the heat exchanger was restored. The developed formulation is believed to be most effective for all heat exchangers used in water application. Keywords- Accelerator, Low Conductivity Water Plant, Heat Transfer, Chemical Cleaning 46 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-125 NITROGEN COMPOUNDS FORMATION IN N2-WATER AND N2-MOISTURE SYSTEMS G.R. Dey# and T.N. Das Radiation & Photochemistry Division, Bhabha Atomic Research Centre Trombay, Mumbai 400085 # Corresponding author: [email protected] Abstract The generation of nitrogen compounds such as NO, NO2, NO2- and NO3- in aqueous and gas phase present in a high ionizing radiation zone is normal phenomena. Their formation mechanisms, and the control processes still pose a challenge with reference to the resulting corrosive environment and its effect on the structural materials used in nuclear industry. The source of nitrogen for these products is mainly from air ingress to the system, and/or the nitrogen compounds such as amines mostly used to control dissolved oxygen. These amines such as ammonia, hydrazine, volatile amines are used in different parts of the nuclear power plants for different purposes viz. pH control, and dissolved oxygen scavenger in coolant or moderator systems. During their uses under high radiation environment N2 and these nitrogencontaining compounds receive low to high doses that affect the compounds’ subsequent chemistry and possibly generate these N-O compounds. With this concern our objective was to study the radiation chemical effects of nitrogen (air ingress), and more recently in nitrogen–moisture cold plasma system. Cold plasma is an electric discharge in presence of dielectric surface(s) such as pyrex, quartz and alumina to produce excited state species, charged species, free radicals and photons near room temperature and atmospheric pressure. Moreover, in N2-moisture cold plasma systems, ozone was not observed as product whereas the absorbance at 204, 214, 226 and 400 nm were observed in gas phase UV-vis spectrophotometric measurements due to NO2 formation. In this presentation a summary of the results on various aspect of the formation of different N-O compounds during radiolysis of aqueous systems as well as in gas phase cold plasma will be discussed. 47 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-126 EVALUATION OF ALUMINUM BRASS COUPONS IN BWR CONDENSATE ENVIRONMENT IN PRESENCE OF METAL IONS K K Bairwa1, V S Tripathi1#, A Kumar2 and D B NaiK1 1: Radiation & Photochemistry Division, Bhabha Atomic Research Centre, Mumbai400085, INDIA 2: Chemistry Division, Bhabha Atomic Research Centre, Mumbai-40085, INDIA # Corresponding Author: [email protected] Abstract Effect of cobalt and cesium ions in the simulated BWR condensate environment (two phase water at 150 oC) on the oxide formed on the aluminium brass has been studied by exposing active and prepassivated coupons in respective environments. Surface changes in the exposed coupons were evaluated by SEM, EDAX and electrochemical studies. The SEM and EDAX data of the exposed coupons indicated marked difference in the surface morphology with varying water chemistry. Presence of nodular grains were seen on SEM images of the pre-passivated Al brass coupons in the Co based media while more granular oxide formation could be seen in presence of Cs. With the mixture of Co and Cs, oxides with larger particle size were seen in the SEM images. The weight change measurement also indicated that Co affects the outer oxide layer to a higher extent as compared to Cs. EDAX measurements indicated incorporation of Co in the oxide layer for the coupons exposed in the Co based media whereas higher aluminum composition was seen in the oxide layer for the coupons exposed in the Cs based media. Cathodic reduction of the oxide layer in sodium perchlorate medium indicated that the oxide grown in only water based media are primarily Cu2O with a minor amount of ZnO but there is a significant amount of Co in the oxide layer for the coupons exposed in the Co based medium. Impedance measurement of the coupons indicated similar protective nature of the passive layers formed under various conditions based on the values of charge transfer resistance obtained by fitting the experimental data to the Randles circuit. However, the higher capacitance values for oxides formed in the Co based medium indicated its porous nature. Thus, there is significant sorption of Co in the passive layer of the aluminium brass while there is no evidence of Cs sorption over aluminium brass could be obtained in the present study. 48 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-127 SYNTHESIS AND CHARACTERIZATION OF V(HCOO)2·2H2O V S Tripathi1#, K K Bairwa1, S N Achary2 and D B NaiK1 1: Radiation & Photochemistry Division, Bhabha Atomic Research Centre, Mumbai400085, INDIA 2: Chemistry Division, Bhabha Atomic Research Centre, Mumbai-40085, INDIA # Corresponding Author: [email protected] Abstract Decontamination of light water reactors wherein stainless steel is used as major structural material involves removal of various substituted ferrites with high lattice energies. Strong reducing agents such as V (II) and Cr (II) are known to be very effective in reductive dissolution of such oxides. The stringent requirement of inert condition poses immense handling related issues for the large scale application of these formulations which can be best overcome by application of their solid compounds. Solid V(II) formate has been synthesised and characterized in this regard. In situ generated vanadyl formate has been reduced with Zn amalgam at high concentration (350 mM). The brown precipitate obtained by the chemical reduction has been found to be a V(II) compound with partial solubility in water. However this compound could be dissolved in equimolar formic acid solution. The UV visible spectra of the the re-dissolved compound indicated the presence of V(II) species. The chemical formula of the compound was established by combination of compositional analysis, FT-IR, thermogravimetry and XPS analysis to be V(HCOO)2·2H2O. Reducing a V(V) precursor led to zinc substituted product formation which could be avoided by using the V(IV) precursor. The phase analysis of the compound was done to get some insight into its structures. The observed XRD data was indexed and the unit cell parameters thus obtained were refined by the least square method. Monoclinic unit cell parameters obtained for the V(II) compound have similarity with the various reported M(II) formate di-hydrate (M = Cu, Zn, Fe). 49 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-129 STUDIES ON FAILURE ANALYSIS OF STAINLESS STEEL ION EXCHANGE HOPPER AT NAPS Ranjana Kusari, S.K.Upadhyay# and Brij Mohan Narora Atomic Power Station Nuclear Power Corporation of India Limited # Corresponding Author: [email protected] Abstract Moderator and PHT purification system is designed to remove the ionic impurities as well as radioisotopes from the Moderator & PHT system. The ion exchange hopper contains deuterated nuclear grade resin (MB H+ for Moderator and MB Li for PHT system) filled in stainless steel hoppers. Same column is used either in Moderator or in PHT system based on the type of resin charged. Stainless steel ion exchange hoppers made of SS plate (SS304L) with 6mm thickness and capacity 135 litres are used in Moderator and Primary Heat Transport system ion exchangers. These Stainless Steel hoppers were being used for the past 24 years. Gradual increase in tritium DAC was observed in the purification building. The ion exchange column was isolated one by one for identification of leakage. DAC has gradually decreased after isolation of the ion exchange column at the leaky pits. The suspected leaky columns which were isolated were removed from location, then these columns were decontaminated visual inspection reveals pitting mark on the column. A portion of the most leaky hopper SSH#10 was cut for investigation. Stereoscopic observations, revealed the type of cracks which was transgranular in nature. The cracks were observed to propagate from outer surface to inner surface. A cut portion of hopper#10 was sent to BARC for further investigation. Upon thorough studies, it was revealed that the failure might be due to iron embedded stress corrosion cracking. This is a generic problem, as found in several ion exchange column. 50 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-131 ANTIMONY (Sb) SORPTION AT HIGH TEMPERATURE AND PRESSURE ON ZIRCALOY, CARBON STEEL (CS) AND MAGNETITE COATED CS (MCS) SURFACES S. J. Keny1, B. K. Gokhale1, A. G. Kumbhar1#, Santanu Bera2, Saibal Basu3, S. Velmurugan2 RPCD1, WSCD2, SSPD3 Bhabha Atomic Research Centre, Trombay, Mumbai – 400 085 (INDIA) * Corresponding author: [email protected] Abstract In Pressurized Heavy Water Reactors (PHWRs), due to aggressive conditions Sb from PHT (primary heat transfer) pump bearings, releases into the reactor core and gets activated to 121Sb and 123Sb. Subsequently, it deposits on out of core surface resulting in radiation exposure to station personnel’s and apparent high decontamination factors. Sb, thus deposited can’t be removed by normal decontamination process. To simulate reactor conditions CS/ MCS/ zircoloy coupons along with Sb metal were exposed to pH 10.20 solution (LiOH) at 280˚C (70-75 Kg/Cm2) in a Teflon coated static autoclave for 30 days. The GIXRD of zircaloy exposed coupons showed the peak position at 2θ ~28.4, 40.5 and 45.5, specifies the presence of Sb as Sb2O4 creating a separate phase on the surface. Whereas weak signal of Sb oxide formation were observed on CS and Magnetite CS surfaces. Released of Sb into the solution in the presence of Zircaloy was found to be maximum (57 ppm) as compared to the CS (28 ppm) and magnetite CS (17 ppm) coupons in the solution. XPS studies indicated very less (2%) incorporation of Sb in CS compared zircaloy (9%) and MCS (14%). This shows that on bare CS surface Sb may not be finding the way due to competition with magnetite formation. Whereas on already formed magnetite major part goes as Fe2+ substitution and on zircaloy it deposits as separate Sb2O4 phase. In all the cases, binding energy of Sb 3d3/2 was found in the range 539.7 – 540.1 eV which indicates the presence of Sb in 3+ state. This study may be useful in developing a formulation for Sb decontamination and reduce Man Rem expenditure in nuclear power plants. 51 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-132 EFFECT OF ANTIMONY(III) ON CARBON STEEL CORROSION INHIBITION BY MOLYBDATE IN CITRIC ACID SOLUTION Vinit K. Mittal, Y. Raghavendra, Santanu Bera, S. Sumathi, S. Rangarajan, S.V. Narasimhan and S. Velmurugan Water and Steam Chemistry Division BARC Facilities, Kalpakkam 603102, Tamil Nadu Abstract: Molybdate is known as a good corrosion inhibitor of carbon steel (CS). But it cannot inhibit CS corrosion in citric acid solution at 85oC. It has been observed that the presence of small concentration of Sb(III) along with MoO42- inhibits CS corrosion efficiently. The corrosion inhibition by MoO42- have been studied extensively by varying the concentration of Sb(III) and MoO42-. A critical concentration of MoO42- is required to passivate CS in acid medium in the presence of Sb(III). The study shows that molybdate forms a thin protective layer on CS surface in presence of Sb(III) which provides the corrosion inhibition. Inhibition property and the layer composition on CS surface have been studied by electrochemical and surface analytical techniques. The protective layer is found to be composed of both Mo and Sb and appears to be formed due to cathodic reduction of Mo6+ to Mo5+ & Mo4+ and anodic oxidation of Fe and Sb. 52 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-134 RADIOACTIVE LIQUID WASTE DISCHARGE REDUCTION STRATEGIES AT TAPS 1&2 Deepa Papachan#, A.K.Panda, S.M.Maskey, M.Joshi, V.S.Daniel Technical Services Section, Tarapur Atomic Power Station 1& 2 # Corresponding Author: [email protected] Abstract Tarapur Atomic Power Station -1&2 (TAPS-1&2) consists of twin unit of Boiling Water Reactors (BWR) and is located at Tarapur in India. The radioactive effluent release from the station is regulated by Atomic Energy Regulatory Board(AERB) in India. Based on the decreasing trend of radioactive liquid waste discharges over a period of a decade, which was within the previously stipulated limits so as to restrict the outside population to receive a radiation dose less than1mSv/year from all type of radioactive releases, it has further brought down the discharge limits. Over the period TAPS1&2 has reviewed the pattern or source of liquid waste generation and the waste treatment processes incorporated at the Station and at Waste management facility (TRAP) affiliated to the station and has been able to bring down the liquid waste discharges to as low as possible within the available infrastructure and financial constrictions. This paper discusses the evolution in the liquid waste management strategies at TAPS1&2 for the past eight years and compares with the international trends and practices to chart out the future course of action. 53 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-135 EVALUATION OF ADVANCED HOT CONDITIONING PROCESS FOR PHWRS P.Chandramohan#, M.P.Srinivasan and S.Velmurugan Radiation and deposit control studies section, Water and steam chemistry division, Chemistry group BARC, Kalpakkam Tamilnadu-603102 # Corresponding email: [email protected], [email protected] Abstract Hot-conditioning/hot functional test process is carried out to the PHT system of reactor before reactor going to critical/operational. The process is aimed in checking the component functionalities at high temperature and high pressure conditions, the process also checks/ removes the suspended corrosion products in heat transport circuit. This process leads to formation of a passive or corrosion oxide film on the heat transport circuit surfaces which protects/mitigates the corrosion of the system circuits during the operation of plant. Major concerned alloy in the Primary Heat Transport (PHT) system of Indian PHWRs during the hot conditioning process and also during operation is the carbon steel due to its high corrosion. Hot-conditioning process mitigates the corrosion of carbon steel by the formation of iron oxide (Fe3O4) as major oxide phase layer on the carbon steel surface with a typical thickness of 1.0 µm with particle size of 1µm after 336h of process at 250 C. But this passive oxide film thickness increase with time of operation of system with c.a. 10µm for 2.2 EFYP. The protectiveness of passive layer can be further enhanced by reducing the particle sizes in the passive film to nano meter range. The process can impact on the compactness of passive oxide layer with reduced pores in the oxide layer and properties of the nano nature oxide (transport properties) impacting the corrosion mitigation. The corrosion mitigation reduce the source term in the activated corrosion product generation. To achieve this a new process ‘Advanced hot conditioning’ was developed in water steam chemistry division, BARC for getting a passive oxide film with a lowered particle size in the passive film. The AHC process with 1g/L of PEG-8000 at 250 C for 336 h showed a particle size 100nm. The process was tested under the normal operating conditions as function of the time, the corrosion parameter like oxide film thickness, corrosion rate and metal ion release to the solution as corrosion products showed improved results for the AHC process. The studies also focused applicability of the above process at high temperature like 260 and 280 C. 54 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-137 . TREATMENT OF FAST REACTOR LIQUID WASTE- ELECTROCHEMICAL METHOD 1 Swapan Kumar Mahato , R. Sudha1#, P. Muralidaran2 and S. Anthonysamy1 1 Chemistry Group, 2 Reactor Operations and Maintenance Group Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, Tamil Nadu, India #Corresponding author: [email protected] Abstract During the operation of fast reactors, components get wetted by sodium. The sodium wetted primary components such as pumps and intermediate heat exchangers (IHX) in fast reactors are cleaned free of sodium followed by suitable chemical decontamination process before taking them for maintenance or for disposal. This helps in reduction of radiation dose to the operating personnel. Sodium cleaning and decontamination generates large volumes of liquid effluent. The major activity in the liquid effluent during sodium cleaning/decontamination is de to Na-22, Mn-54, Co-58, Co-60, Fe-59, Cs-137 and Cs-134. It is required to chemically treat the effluent to reduce the activity levels prior to storage in tanks and transportation to the waste management facility for final disposal. Conventionally the ion exchange method is used for removal of radionuclides which produces large quantities of secondary waste. A method which is suitable both for removal of radionuclides present in low concentration and that avoids generation of large quantities of secondary waste is required. Hence an electrochemical method for metal ion removal is attempted in this work which produces little or no secondary waste. Electrochemical method towards removal of manganese ions was finalized earlier using reticulated vitreous carbon (RVC) from simulated decontamination solution containing a mixture of sulphuric and phosphoric acids. In continuation of the experiments for the removal of cesium ions from simulated cleaning solution which has an alkaline pH, a thin film of nickel hexacyanoferrate (NiHCF) was deposited electrochemically on the surface of RVC. Hexacyanoferrates are known for selectively binding cesium. This NiHCF coated RVC was used for electrodeposition of Cs ions. NiHCF coated and Cs deposited RVC was characterized using SEM/EDS analysis. EDS analysis confirms the presence of Cs on NiHCF coated RVC. 55 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-138 FIXATION OF NUCLEAR WASTE INTO GLASS MATRICES FOR ULTIMATE DISPOSAL G. Hazra1#, T Das1 and P. Mitra2 1 Nuclear and Analytical Chemistry Section, Department of Chemistry,The University of Burdwan, Burdwan-713409, W. B., India. 2 Department of Physics, The University of Burdwan, Burdwan-713409, W. B., India. # Corresponding Author: [email protected] Abstract For the long-term storage of high-level nuclear wastes (HLW), it is required to develop glasses with high chemical durability, thermal stability and waste solubility. In comparison with high silica and borosilicate glasses, lead-iron phosphate glasses have the advantages of good waste solubility and low melting temperatures so that volatile hazardous radionuclides (e.g. Ru, Tc and Cs) can also be incorporated in the glasses. Lead–iron phosphate glasses were proposed as the potential nuclear waste glasses, which show more durability than that of a comparable borosilicate waste glass. In the present works simulated glass (LIP) composition as that of the waste are selected and melted from batch oxides. Finally these glasses after setting proper melting conditions are subjected to leaching studies under Soxhlet condition at different medium. The melting point of the LIP4 melted with PbO as source was found to be comparatively lower (~750°C) than in case of those melted with Pb3O4 (> 900°C). The pH increase of the leachate solution can be distinguished and correlated to the schematic phosphate network. At the beginning the P–O–M (M= Pb, Sr, Ce, Ba) bonds could be preferentially corroded with a release of phosphate groups in the leaching solution contributing to make up a drop in pH. Later on, the acidic medium could involve a corrosion enhancement with partial hydrolysis of P–O–P band P–O–Pb bonds. As the glass is quenched from the melt, the Fe-O-P-O-Pb network formed with voids that can be occupied by waste ions such as U4+. The leaching rate of the LIP glasses (both with and without U) containing ceria are lower than without ceria. This is due to having higher coordination number linked with oxygen in the phosphate network. Hence the durability of the LIP glasses may be increased with the addition of CeO2. 56 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-139 ANTIMONY SORPTION PROPERTIES OF CHITOSAN – NANO TIO2 COMPOSITE BEADS Padala Abdul Nishad, Anupkumar Bhaskarapillai, Sankaralingam Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities, Kalpakkam, Kancheepuram, Tamil Nadu – 603102 #Corresponding Author: [email protected] Abstract Routine decontamination campaigns of nuclear reactors are generally effective in removing various radionuclides such as cobalt, caesium, etc., and bring down the radiation field. However, during some of the decontamination campaigns, the radiation field at some surfaces were seen to have actually gone up. This was found to be due to lack of removal of antimony isotopes by the regular ion exchange resins used, which subsequently deposited over out of core surfaces leading to increased radiation field on those surfaces. Thus there exists a need for efficient antimony removal system. We have synthesised nano titania impregnated - epichlorohydrin crosslinked chitosan beads, which were found to have high sorption capacity for antimony. The beads, which were synthesised in formats suitable for large scale (column mode) applications, were shown to be effective sorbent of antimony in both +3 and +5 oxidation states. The sorbent exhibited complete removal of antimony from its aqueous solutions of concentration ranging from 150 ppb to 120 ppm. In order to understand the sorption mechanism and to fine tune the bead composition, the effect of crosslinker concentration used during the synthesis on the swelling and sorption properties of the beads was investigated in detail. The variation effected significant changes in physical parameters such as bead diameter, swelling ratio, equilibrium water content and true wet density. Sorption capacity, unlike with regular resins, was found to increase with increase in crosslinker amount. The antimony sorption capacity of the crosslinked beads prepared by crosslinking 0.3 g uncrosslinked beads with 6.4 mmol epichlorohydrin (crosslinker) was 493 µmol/g. Noncrosslinked beads showed a capacity of 75 µmol/g, while the crosslinked beads made with the least amount of crosslinker (0.64 mmol per 0.3 g beads) showed a capacity of 133 µmol/g. These results indicate the possible involvement of the crosslinker in the sorption. 57 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-141 HEAVY METALS-BIOREMEDIATION BY HIGHLY RADIORESISTANT DEINOCOCCUS RADIODURANS BIOFILM PROSPECTIVE USE IN NUCLEAR REACTOR DECONTAMINATION Sudhir K. Shukla, T. Subba Rao# Biofouling & Biofilm Processes Section, Water and Steam Chemistry Division, BARC facilities, Kalpakkam, 603 102 India, # Corresponding author: [email protected] Abstract Over the past few decades, rapid growth of chemical industries has enhanced the heavy metal contamination in water, thereby raising environmental concerns. In the nuclear power industry, decontamination procedure also generates radioactive heavy metal containing wastes. Radio-resistant Deinococcus radiodurans R1 is reported to be a potential candidate for the treatment of low active waste material. To use any bacterium for bioremediation purpose, knowledge about its biofilm production characteristics is a prerequisite. This is because biofilm-mediated bioremediation processes are more efficient as compared to processes mediated by their planktonic counterparts. However, so far there are no reports on the biofilm producing capability of D. radiodurans. We observed that tagging of D. radiodurans by a plasmid harbouring gfp and kanR conferred significant biofilm producing property to the bacterium. Chemical analysis of biofilm matrix components produced by D. radiodurans showed that the matrix consists primarily of proteins and carbohydrates with small amount of extracellular DNA (eDNA). Further, we studied the effect of Ca2+ on D. radiodurans biofilm formation and it was observed that D. radiodurans biofilm formation was enhanced at higher concentrations of Ca2+. We investigated the capability of D. radiodurans biofilm to remove the heavy metals Co and Ni from synthetic waste streams. Results showed that Ca2+ enhanced the bioremediation of both heavy metals (Co, Ni) by D. radiodurans biofilms in a highly significant manner. In the presence of 50 mM Ca2+ 35% Co removal and 25% Ni removal was observed, when compared to biofilm grown in the absence of Ca2+, which showed mere 7% Co and 3% Ni removal, respectively. The results showed that the presence of Ca2+ significantly enhanced exopolysaccharide and eDNA (both negatively charged) production in the biofilm matrix. This indicated adsorption could be the major mechanism behind enhanced biofilm mediated removal of heavy metals. The study signifies the potential use of D. radiodurans biofilms, which can tolerate >20 kGy in nuclear reactor decontamination process for the removal of active heavy metals. 58 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-142 OPERATING CONDITIONS INFLUENCE CORROSION OF CARBON STEEL IN A FRESHWATER DISTRIBUTION SYSTEM T. Subba Rao Biofouling & Biofilm Processes Section, Water & Steam Chemistry Division, BARC Facilities, Kalpakkam, 603 102 India # Corresponding author: [email protected] Abstract The influence of operating conditions (flow and no flow situations) on the corrosion of carbon steel(CS) were simulated and investigated. Conventional microbial culture methods and molecular tools were used to characterize the biofilm and corrosion causing bacteria. Denaturing gradient gel electrophoresis showed significant diversity and variation in the bacterial community. Raman spectroscopy was used to characterize the corrosion deposits, the following iron oxide phases were identified; lepidocrocite, goethite, hematite and magnetite. Transformation of two iron oxides hematite and magnetite vice versa was noticed in the experimental system. In conclusion a plausible CS corrosion control method was described. 59 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-143 ISOLATION AND CHARACTERIZATION OF THE MICROBIAL COMMUNITY OF A FRESHWATER DISTRIBUTION SYSTEM P. Balamurugan1, T. Subba Rao# Biofouling & Biofilm Processes Section, Water & Steam Chemistry Division, BARC Facilities, Kalpakkam, 603 102 India 1 Dept of Biotechnology, Pondicherry University, Puducherry, India # Corresponding author: [email protected] Abstract This investigation provides generic information on culturable and non-culturable microbial community of a freshwater distribution system. Culture based and culture independent (16S rRNA gene sequencing) techniques were used to identify the resident microbial community of the system. Selective isolation of the fouling bacteria such as biofilm formers and corrosion causing bacteria was also attempted. Denaturing gradient gel electrophoresis (DGGE) was carried out and the bands were sequenced to obtain the diversity of the total bacterial types. Pseudomonas aeruginosa was predominantly observed in most of the samples. A variety of bacteria, related to groups such as Cyanobacteria, Proteobacteria, Actinobacteria, Bacteroidetes and Firmicutes were identified. The study highlights the relevance of the observed microbial diversity with respect to material deterioration in a freshwater distribution system, which can aid in designing effective control methods. 60 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-144 MICROFOULING ASSESSMENT AND ITS CONTROL IN A HEAVY WATER PRODUCTION UNIT Rajesh Kumar, T. Subba Rao# Biofouling & Biofilm Processes Section, Water & Steam Chemistry Division, BARC Facilities, Kalpakkam, 603 102 India # Corresponding author: [email protected] Abstract The water treatment plant (WTP) of a heavy water production unit was extensively fouled by microorganisms. On-site investigations showed severe algal and bacterial growth in the various units of WTP and very dense microbial fouling in the vacuum degasser (VD) unit. Digital and microscopic images showed that the microfouling problem was primarily due to a slime bacterium and a fungus. Microbiological analysis showed a bacterial count of ~105 cfu ml-1 in the various sections of WTP. The slime/biofilm scrapings had very high bacterial population (>109 cfu cm-2). High organic carbon values in the system (5.0 to 19.5 ppm) had supported microbial growth in WTP and augmented resin fouling. Chlorination was inadequate in controlling microfouling, consequently chlorine dioxide was tested and found to be a better biocide. A 2.0% sodium omadine solution had completely inhibited the fouling fungus. 61 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-145 CORROSION OF ALLOY D9 IN LIQUID SODIUM R. Sudha1#, K. Chandran1, P. Muralidaran2 and S. Anthonysamy1 1 Chemistry Group, 2 Reactor Operations and Maintenance Group, Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, Tamil Nadu, India # Corresponding author: [email protected] Abstract Alloy D9 (15Cr-15Ni-Mo-Ti-Si) is chosen as clad and wrapper material for 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. The successful operation of the reactor depends on the compatibility of the core structural materials such as clad and wrapper which are subject to high neutron irradiation and in contact with high temperature liquid sodium. The chemical compatibility of sodium with the core structural materials is generally good when the sodium is in the pure state. One of the major criteria for selection of clad and wrapper materials is corrosion in liquid sodium environment. The corrosion of alloy D9 in presence of sodium was studied at different temperatures (773, 798, 823 and 873 K) for various duration of time ranging from 1000 to 3000 h in a well characterized sodium loop. The observed corrosion data such as weight loss, depleted layer formation, changes in microstructure of the alloy D9 specimens are compared with the literature data available on sodium corrosion in static condition. The corrosion rates in dynamic and static sodium are comparable in the measured temperatures as the purity of sodium with respect of oxygen concentration is comparable. The corrosion rate is faster at higher temperature and for longer duration of exposure to liquid sodium. Loss of thickness of material at 873 K is less than 5 µm per year. Exposure to sodium causes selective dissolution of alloying elements such as nickel and chromium. The metal losses, thickness of corroded layers and changes in chemical composition were only marginal at a maximum temperature of 873 K. Hence alloy D9 is a good choice for clad and wrapper tube material. 62 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-146 THREE DECADES OF EXPERIENCE WITH COOLING WATER SYSTEM OF A FAST REACTOR A.Suriyanarayanan# and B.S.Panigrahi Reactor Chemistry Section, Reactor Operation & Maintenance Group, Fast Breeder Test Reactor, Indira Gandhi Centre for Atomic Research, Kalpakkam - 603102 (TN), India # Corresponding Author: [email protected] Abstract The cooling water system constitutes the terminal heat exchange system for the fast breeder test reactor (FBTR) which is a sodium cooled fast reactor of 40 MWt capacity. It transfers the residual heat to atmosphere through a cooling tower. Cooling water system of FBTR comprises two sub-systems namely condenser cooling water system and service water system. Condenser cooling water is circulated through main condenser, dump condenser, condensate cooler, generator air cooler and turbine oil cooler. Service water system removes heat from several heat exchangers of auxiliary systems like air compressor, cold trap cooling, nitrogen plant, Biological Shield Cooling (BSC), Diesel Generator (DG) and steam-water system sample coolers. The cooling water system consists of an open recirculating type with an induced draft cooling tower as the ultimate heat sink. Initially, Palar river water was used as the cooling medium. At present, due to scarcity of river water, sub soil water and output from Nuclear Desalination Demonstration Plant (NDDP) are also used as cooling water. The material of construction of pipe line is carbon steel and the heat exchanger tube and other equipment materials are copper, admiralty brass, aluminium brass, bronze, Cu-Ni and carbon steel. The construction of the cooling water system of FBTR was completed in 1980. Since then the sub-systems were commissioned one by one. Whenever a sub system was commissioned, it generated a lot of impurities which affected the existing treatment programme. Sodium hexa meta phosphate treatment, Langelier Index monitoring, chlorination, global and target dispersant addition at high heat flux heat exchanger , chemical cleaning of corroded pipelines, corrosion monitoring , side stream filtration, addition of phosphonate-based corrosion inhibitor, broad spectrum biocide and specific biocide for iron oxidising bacteria are some of the phases of the cooling water treatment programme. At present, corrosion rates are generally less than 3 mpy for carbon steel and less than 0.5 mpy for brass. This paper details the challenges faced and remedial measures implemented in the cooling water system of FBTR for better performance and increased availability. 63 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-156 WATER TREATMENT WITH CHLORINE: INFLUENCE OF SOURCE WATER CHARACTERISTICS ON CHLORINATION & CBPS FORMATION R K Padhia, S Subramaniana, K K Satpathya# a Environment & Safety Division, RSEG /EIRSG, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu- 603 102, India. # Corresponding author: [email protected] Abstract Chlorine is cheap, reliable and proven oxidant used worldwide for drinking water disinfection and bio-fouling control in water utilities. Formations of toxic trihalomethanes (THMs) due to the reaction of chlorine with natural organic matter present in water have raised concern over its use for water treatment. Ambient source water characteristics and operational parameters dictate the THMs load in the chlorinated water. To evaluate the effect of change in water quality parameters on chlorination and chlorination byproducts (CBPs) formation, laboratory chlorination experiments were carried out for two source water viz. Palar River (PR), Open reservoir (OR). Chlorine demand (CD) and total trihalomethanes formation potential (TTHMF) were measured for different contact time. CD and TTHMF were always observed to be higher for open reservoir water. Chlorine demand value for open reservoir ranged from 1.48 to 2.43 mg/L and that of Palar water ranged from 1.01 to 1.76 mg/L. TTHMF potential (TTHMFP) of Open reservoir ranged between 61-98 µg/L which was relatively higher compared to that for Palar (38-94 µg/L). Water quality descriptors such as temperature, pH, Chlorophyll and dissolved oxygen of subsoil Palar River water was found to undergo substantial alteration upon open storage in the open reservoir. Dissolved organic content (DOC) of the two water sources studied i.e Palar river subsoil water (PR) and open reservoir water (OR) varied from 0.41 to 0.95 mg/L and 0.93 to 2.53 mg/L respectively. Presence of bromide in these water sources (0.15 – 0.26 mg/L in PR and 0.17 -0.65 mg/L in OR) have resulted significant brominated THMs. The formation of more amounts of brominated THMs lead to enhanced toxicity load in the chlorinated water. Key words: Chlorination, Chlorination byproduct, Chlorine demand, Dissolved organic content Trihalomethanes 64 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-158 ENTRAINMENT AND IMPINGEMENT OF AQUATIC FAUNA AT COOLING WATER SYSTEM OF MADRAS ATOMIC POWER STATION (MAPS) S. Barath Kumar*, N. P. I. Das and K.K. Satpathy Environmental and Safety Division, Radiological Safety & Environmental Group, Electronics Instrumentation & Radiological Safety Group, IGCAR, Kalpakkam, Tamil Nadu, India-60310 *Corresponding author: [email protected] Abstract In order to understand the entrainment and impingement of marine fauna at the cooling water system of Madras Atomic Power Station (MAPS), a pilot study programme for one year (March 2013 to February 2014) period was carried out. In this regard entrapped fauna were recorded on hourly scale at three cooling water screens of MAPS on weekly basis (76 sampling). The entrained specimens were categorized, weighed and observed for impingement effect. The study showed that the major entrained groups of animals were jellyfish, crab, fish and shrimp. Apart from above, a few cephalopods and sea snakes were also observed as entrained entity. Totally 67 species of marine faunas impinged on the water intake screens of MAPS during the study. The numerical count of the total observations during the study period showed jellyfishes were the largest entrained group covering around 44.85% of individual and constituting almost 94.82 % of biomass recorded during the study period and sea nettle jelly (Chrysaora quinquecirrha) was impinged with highest frequency. The monthly data analysis showed large scale jellyfish entrainment during March, April, August and October. This is mostly due to the jellyfish bloom at Kalpakkam coast during this period. The next entrained group was crab, which counts 30.37% individuals and on biomass counts for 2.93% of the total entrained fauna. A total of 16 species of crabs were observed as entrained groups, of which Swimming crab (Charybdis lucifera), Stone crab (Menippe rumphii) species were frequently impinged (> 63% occurrence). Both the crab species have no or low commercial value as found from the local fisher folk. The next entrained group was fish which accounts for 14.84 % of individual count and mere 1.67 % of biomass. Totally 33 number of fish species were observed. The highest impinged species were pony fishes (Secutor ruconius, Secutor insidiator, Photopectoralis bindus, Alepes kleinii and Leiognathus equulus) (21% occurrence). These few entrained fishes are mostly very small in size and have less commercial value. The other entrained group such as shrimps (0.31%), sea snakes (0.19 %) and cephalopods (0.06%) were rarely observed and negligible in biomass count, comprising less than 1% of entrained biomass. The monthly study trend showed crab impingement declines from September to February. In case of fish, the highest impingement was observed during the month of December, on night and on full moon day. In the total faunal highest impingements occurred at night time, full moon day and low tide compared to the day time, new moon day and high tide. The present data when compared with the impingement data from other coastal power plants, shows that the impinged fish biomass at MAPS cooling water system is much less than the other temperate and tropical power plants. 65 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-159 SURFACE AND ELECTROCHEMICAL CHARACTERIZATION OF NANO ZINC FERRITE COATING ON CARBON STEEL Sumathi Suresh, S. Rangarajan# and S. Velmurugan Water and Steam Chemistry Division BARCF, Kalpakkam, INDIA # Corresponding author: [email protected] Abstract The structural materials in the nuclear power reactors are mainly iron and nickel based alloys. Operation of these nuclear reactors at high temperatures and high pressures for a longer duration leads to the formation of various oxides due to the corrosion of the structural materials and the nature of these oxides depend on the chemical environment prevailed. Since the corrosion process is usually electrochemical in nature, the interface formed between the alloys and the oxides play a crucial role in deciding the overall corrosion resistance of the structural materials. Therefore, modifying these oxides to nano size would improve the adherence and protectiveness of the interfacial film. In this context, the chemical synthesis of zinc ferrite (ZnFe2O4) was carried out by precipitation method using zinc sulphate and iron ammonium sulphate. The synthesized ferrite powder was confirmed by Raman Spectroscopy. X-Ray Diffraction studies showed that the intensity and the ‘d’ values of the entire observed diffraction peaks perfectly match with the single-crystalline cubic spinel form of zinc ferrite having lattice constant a= 8.436 Å. The ferrite targets were prepared (10 mm diameter pellet with a calculated density of 4.22 gm/cm3) using synthesized ZnFe2O4 powder by sintering at 1000°C for 24 hours. Thin film of ZnFe2O4 was deposited on Carbon Steel specimens using pulsed laser deposition technique. Characterization of the deposited ferrite was carried out using Laser Raman, X-Ray Diffraction, X-ray Photoelectron Spectroscopy and Secondary Electron Microscopy. Raman data of the coated ZnFe2O4 matched with the standard ZnFe2O4 oxide. X-ray diffraction pattern indicated that the sample was in single phase with an average grain size 30 nm. XPS data clearly indicated the formation of the ZnFe2O4. Scanning electron microscopy and atomic force microscopy techniques were used to analyze the film surface morphology. The mechanism of corrosion resistance / improvement in the deposited layer was studied by electrochemical techniques and the results are presented in detail in this paper. 66 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-163 EVALUATION OF CORROSION INHIBITORS FOR HIGH TEMPERATURE DECONTAMINATION APPLICATIONS V. S. Sathyaseelan, A. L. Rufus and S. Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities, Kalpakkam, Tamilnadu – 603 102, INDIA # Corresponding email: [email protected] Abstract Normally, chemical decontamination of coolant systems of nuclear power reactors is carried out at temperatures less than 90 °C. At these temperatures, though magnetite dissolves effectively, the rate of dissolution of chromium and nickel containing oxides formed over stainless steel and other non-carbon steel coolant system surfaces is not that appreciable. A high temperature dissolution process using 5 mM NTA at 160 °C developed earlier by us was very effective in dissolving the oxides such as ferrites and chromites. However, the corrosion of structural materials such as carbon steel and stainless steel also increased beyond the acceptable limits at elevated temperatures. Hence, the control of base metal corrosion during the high temperature decontamination process is very important. In view of this, it was felt essential to investigate and develop a suitable inhibitor to reduce the corrosion that can take place on coolant structural material surfaces during the high temperature decontamination applications with weak organic acids. Three commercial inhibitors viz., Philmplus 5K655, Prosel PC 2116 and Ferroqest were evaluated at ambient and at160 °C temperature in NTA formulation. Preliminary evaluation of these corrosion inhibitors carried out using electrochemical techniques showed maximum corrosion inhibition efficiency for Philmplus. Hence, it was used for high temperature applications. A concentration of 500 ppm was found to be optimum at 160 oC and at this concentration it showed an inhibition efficiency of 62% for carbon steel. High temperature dissolution of oxides such as Fe3O4 and NiFe2O4, which are relevant to nuclear reactors, was also carried out and the rate of dissolution observed was less in the presence of Philmplus. Studies were also carried out to evaluate hydrazine as a corrosion inhibitor for high temperature applications. The results revealed that for carbon steel inhibition efficiency of hydrazine is comparable to that of Philmplus, while for stainless steel hydrazine is a better choice. 67 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-171 DEVELOPMENT OF LEACHING METHOD FOR THE ANALYSIS OF PALLADIUM CATALYST USED IN THE MODERATOR COVER GAS CIRCUIT OF MAPS BY ICP-OES S. Vijayalakshmi# and S. Annapoorani, Materials Chemistry Division, IGCAR, Kalpakkam # Corresponding author: [email protected] Abstract The radiolysis of heavy water, which is used as Moderator in PHWRs results in the formation of D2 and O2 which strips into the helium cover gas. Considering the safety aspect, there is a technical specification limit for the concentration of D2 in moderator cover gas and the upper limit is 4%v/v. To control the D2 concentration within this limit during reactor operation, part of the cover gas flow is passed through the recombination units in the moderator cover gas circuit. There are two recombination units and each contains pellets of Palladium coated alumina as catalyst in which the recombination of D2 and O2 gas takes place. To improve the efficiency of the recombination units, catalyst containing 0.5% Pd is preferred over 0.2% Pd and therefore the same was taken up for use in Madras Atomic Power Station (MAPS). Analysis of palladium catalyst was required towards the quality control purpose. The catalyst contains palladium as coating over the alumina pellets. Therefore, in this study, leaching procedure was standardized for the complete removal of palladium and the leached solutions were analyzed by ICP-OES for the determination of palladium. Experimental: Complete dissolution of sample requires fusion. Fusion with 100mg of sample pieces was found to result in the poor precision of around 40%. Considering the difficulty of powdering the sample to get better precision in fusion, leaching procedure was tried out for analysis. 0.5gm of sample was heated with aquaregia till the disappearance of black colour on the alumina balls. The solution was made upto known volume and appropriately diluted for analysis. The complete removal was also checked by repeating the leaching with already leached sample. Concentration calibration using 340.458nm line was employed for the analysis. Both samples (sample containing 0.2% Pd and sample containing 0.5% Pd) were analyzed in duplicate and the results were found to agree well with the expected value. Conclusion: A simple and convenient leaching procedure was standardized and recommended for the analysis of palladium catalyst and it was also applied to the samples from MAPS. Acknowledgement: Authors would like to acknowledge Chemistry Control Section of MAPS for providing catalyst sample for standardizing the method and Dr. K. Sankaran, Head, Analytical Chemistry and Spectroscopy Section, MCD, CG for his support in carrying out the analysis. 68 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-172 DISSOLUTION OF COBALT METAL POWDER V.S. Sathyaseelan, A. L. Rufus and S. Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities Kalpakkam (TN) – 603 102, INDIA # Corresponding author: [email protected] Abstract During the transfer of Self Powered Neutron Detectors (SPNDs), from the core of the nuclear reactor to fuel storage bay, there is every possibility of cobalt powder spillage. In order to decontaminate the surfaces during such occasions, some studies were carried out to dissolve cobalt powder. Various formulations such as potassium permanganate, permanganic acid, nitric acid and nitrilo triacetic acid (NTA) were investigated for their efficiency in dissolving metallic cobalt taken in the form of powder. Investigations revealed that a two step process involving a surface conditioning step by permanganic acid followed by dissolution of the cobalt powder by 10mM nitric acid will efficiently decontaminate the cavity surface. 69 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-174 STUDIES WITH ANTI FOULING COATING ON SEAWATER INTAKE SYSTEM SCREENS OF MAPS N.Sankar, V.S.Santhanam, P.Umapathi, K.Hari Krishna#, D.Rajendran, MAPS, Kalpakkam Dr.P.S.Murthy, and Dr.V.Venugopalan, BBPS, WSCD, Kalpakkam # Corresponding Author: Abstract Biofouling has been a concern for cooling water systems of coastal power plants and the same is being experienced in Madras Atomic Power Station (MAPS). Macro fouling organisms cause major problems for smooth operation and maintenance of the cooling water system. The cooling water intake structures particularly the screens, which act as the barrier for marine organisms to enter into the cooling water system, gets fouled severely in a short period of time. Though chlorination is being done to control biofouling, it is ineffective due to the inward flow of seawater. Severely fouled gates necessitate frequent cleaning and maintenance which involves lifting of heavy structures, laborious manual cleaning and maintenance. In order to find remedial measures for the said concern, studies have been taken up for identification of simple but effective methods in controlling bio fouling. Accordingly studies with Anti Fouling Coating (AFC) applications have been identified and field studies were carried out to review its effectiveness in meeting the given requirement. One of the gates was coated with Anti Fouling coating (AFC) and exposed to sea water and the bio fouling tendency was regularly monitored. It was noted the AFC coated gate was observed to have less bio fouling compared to the in-practice coal tar epoxy coatings. The small quantity of fouling deposits was generally observed to be on the side opposite to the sea water current. The area exposed to sea water currents had relatively less biogrowth. The dislodgement or removal of bio growth could be achieved by gentle pressure or scrapping thus demonstrating its effectiveness in controlling the bio fouling. Studies are also in progress to with Foul release coatings (FRC) to study its effectiveness. 70 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-175 INFLUENCE OF GEOMETRY OF PIPE ON FLOW ACCELERATED CORROSION - A STUDY UNDER NEUTRAL PH CONDITION P.Madasamy, M.Mukunthan, P.Chandramohan, T.V.Krishna Mohan, and S.Velmurugan# Water and Steam Chemistry Division BARC Facilities, Kalpakkam – 603 102 Tamilnadu, INDIA. Andrews Sylvanusa and E.Natarajanb AU-FRG Institute for CAD/CAM , Anna University, Chennai-25 b Institute For Energy Studies, Anna University, Chennai-25 a # Corresponding author: [email protected] Abstract The carbon steel piping material’s degradation due to flow accelerated corrosion (FAC) is one of the problems in nuclear power plant. FAC impacts plant operation and maintenance significantly. Wall thinning of structural materials should be predictable based on combined hydrodynamics analyses and experimental corrosion data. Such predictive tools help to take preventive measures before loss of material becomes a serious issue for plant operation. In order to develop predictive tools, data on the effect of various parameters that control FAC are required. As per existing literature, one of the important parameters that affect FAC is piping configuration (Geometry of flow path). Hence, experiments were carried out to assess the role played by the geometry of the piping in the FAC of carbon steel. In this study, experiments were conducted in simulation loop under neutral pH condition while varying the geometry parameter of bend such as bend angle and bend radius. Therefore, pipe specimen holder 15 NB bend with 58 o, 73° as bend angle and 4D, 2D bend radius was designed and fabricated. The experiments were carried out in order to quantify the wear rate (wall thickness measurement was by ultrasonic method) with a single phase flow velocity (7 m/s) under neutral pH conditions With the pipe specimen four experiments were conducted under neutral pH condition and at 120 °C. Wall thickness mapping was carried out by ultrasonic thickness gauge using a template before and after the experiment. High wall thickness reduction under neutral water chemistry enables easy measurement by ultrasonic thickness gauge. It was observed from the first two sets (2D58°, 4D58°) that the corrosion rate with 4D, 58° was 50% less than the corrosion with 2D 58o. Subsequently, another two sets of experiments (2D 73o and 4D 73o) was carried out in SIM loop at 7 m/s under neutral pH conditions for two months. Thus, this method of experiments enables us to understand the geometrical influence. The comparison of all the four set of experiments indicated that minimum corrosion rate (1.3mm/year) was obtained with 4D 73 geometry. Further, velocity distribution, wall shear stress in the bend geometry were mapped by Computed Fluid Dynamics (CFD) and correlated with the corresponding measured wear rate. Good correlation was obtained between theoretically obtained corrosion rate and the experimental value. 71 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-176 EVALUATION OF PLASMA COATED CARBON STEEL TO RESIST FLOW ACCELERATED CORROSION P.Madasamy, J. Alphonsaa, J. Ghanshyama, S. Mukherjeea, M.Mukunthan, P.Chandramohan, T.V.KrishnaMohan, ,E.Natarajanc and S.Velmurugan# Water and Steam Chemistry Division BARC Facilities, Kalpakkam – 603 102 Tamilnadu, INDIA. a Institute for Plasma Research, Ahmedabad c Institute For Energy Studies, Anna University, Chennai-25. #Corresponding author: [email protected] Abstract Coatings have historically been developed to provide protection against corrosion and erosion that protects the material from chemical and physical interaction with its environment. Corrosion and wear problems are still of great relevance in a wide range of industrial applications and products as they result in the degradation and eventual failure of components and systems both in the processing, manufacturing and power industries. Various technologies can be used to deposit the appropriate surface protective agents that can resist deterioration under specific conditions. They are usually distinguished by coating thickness as it depends on method of coating. Diffusion based coatings by plasma nitriding have recently been used for improving wear and corrosion resistance properties. A collaborative study on plasma nitriding was initiated with FCIPT, a division of Institute of Plasma Research.This is one of the method to control the wall thickness reduction of carbon steel feeder pipe and the influence of FAC in PHWR. In order to control the influence of Flow Accelerated Corrosion on feeder pipe of PHWR reactor, as a remedy, coating by plasma nitriding process was carried out inside the pipe. The plasma coating is by nitrogen/hydrogen mixture plasma through diffusion process and thickness of coating of upto 300 µm was possible. The coating can withstand a temperature up to 500 °C. The limitation of this process is that it cannot nitride pipes with inside diameter less than 5 mm. Two samples of 15 mm NB Sch 80 straight pipe length of 10 cm pipe module section were plasma nitrided at FCIPT, IPR for optimization of the process parameters. The wall thickness of the sample was measured axially and circumferentially by Ultrasonic thickness gauge with specific marking with templates before carrying out plasma nitriding process. During plasma nitriding, the temperature was maintained at 520 oC for 24 hours. Plasma nitriding process was done using a pulsed DC power supply. We were able to get the required temperature and the results were as per the requirement. There was increase in hardness to 310 HV0.1 from 100HV0.1 at the outer as well as the inner surface with a case depth of more than 250 microns in the inner surface and 650 microns on the outer surface. The samples after coating were checked for thickness variation by Raman spectroscopy as wells as microscopy, and it was found that the coating was uniform and coating consisted of iron nitrides only. The bend specimen optimization of coating parameter is in progress at FCIPT/ Institute of Plasma Research, Gandhinagar. For functional test to check the corrosion resistance, a specimen holder was designed and fabricated for the treated specimen such that it can withstand a velocity of 7 m/s. The holder was mounted in SIM loop in the heater outlet. The SIM loop was maintained at 120 °C and 7 m/s for about 30 days with less than 20 ppb dissolved oxygen condition. The experiment is in progress in SIM loop in order to check resistance to FAC under neutral pH condition. 72 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-177 PREPARATION AND DISSOLUTION OF URANIUM DIBUTYL PHOSPHATE (UDBP) M.K.Dhanesh*, A.L.Rufus and S.Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities Kalpakkam (TN) – 603 102, INDIA *Department of Chemistry, University of Calicut, Calicut, INDIA # Corresponding author: [email protected] Abstract A sticky coating of Uranium Dibutyl Phosphate (U-DBP) was found to be formed on the surfaces of nuclear fuel reprocessing facilities and on the surfaces of reprocessed waste storage tanks. This poses both radiation exposure and criticality hazard. Hence, it is required to periodically dissolve U-DBP deposits. In this connection, an attempt was made to synthesis U-DBP from U3O8 and to evolve a suitable chemical formulation for its dissolution. The U-DBP complex synthesised was yellow coloured sticky deposit. For the dissolution studies, various formulations such as EDTA, disodium-EDTA and sodium carbonate were considered. From the literature it was seen that uranium deposits dissolve effectively in oxidising medium. Hence, dissolution studies were carried out in above formulations in the presence and absence of hydrogen peroxide, which is a well known oxidising agent. From the experimental results it was seen that of all the formulations used, sodium carbonate was found to be very efficient. In the presence of peroxide, the rate of dissolution was found to increase. Investigations were carried out to optimise the concentration of sodium carbonate, peroxide and also the temperature of dissolution. 73 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-178 STUDIES ON GADOLINIUM PRECIPITATION IN MODERATOR SYSTEM OF NUCLEAR REACTOR Akhilesh C Joshi, Puspalata Rajesh, A.L.Rufus and S.Velmurugan# Water & Steam Chemistry Division BARC Facilities, Kalpakkam 603 102, India # Corresponding author: [email protected] Abstract Gadolinium is used in the moderator system of many Pressurised Heavy Water Reactors (PHWRs) for start-up, shut- down and reactivity control during operation. It is very much essential to maintain gadolinium concentration in the system as desired. It has been reported that gadolinium gets precipitated in as oxalate in carbonated water under the influence of -radiation. Hence, studies were carried out to investigate the effect of dose, presence of other metal ions and metal surfaces on the precipitation of gadolinium. The results showed that the amount of carboxylic acids viz., formic acid and oxalic acid, formed due to radiolysis is dependent on the dose. and that the curve passes though a maxima. Gadolinium is added in higher concentration in Advanced Heavy Water Reactor. So, experiments with high concentration of gadolinium were also carried out. Ultra pure water saturated with high purity CO2 containing gadolinium and desired ion/surface was irradiated with γ-radiation from 60Co source at 25oC to doses ranging from 2.5-16.6 Mrad. . At lower doses, formation of carboxylic acids takes place but as the dose increases, decomposition of these acids starts and hence the concentration Vs dose passes through a maximum. It was found that precipitation of gadolinium as oxalate occurred at lower doses. At higher doses, it was seen that pH of the solution decreases and hence solubility of gadolinium oxalate increases. It was also observed that the amount of gadolinium precipitated varied linearly with the initial concentration of gadolinium varying from 2 ppm to 20 ppm. While for gadolinium concentration from 20 ppm to 400 ppm, gadolinium in particulate form was observed. The amount of carboxylic acids formed depends on the nature of cations present in solution. It was found that the amount of oxalic acid formed in the case of gadolinium was more than that formed in the case of sodium. Presence of metal oxides such as ZrO2 formed over zircoloy surfaces was found to enhance the precipitation of gadolinium. This was confirmed from the stronger XPS peak of gadolinium in presence of ZrO2 compared to that in absence of ZrO2 showed the enhanced precipitation of gadolinium in presence of ZrO2.Further, the precipitation of gadolinium was also found to be influenced in the presence of metal surfaces such as zircoloy. 74 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-179 OBSERVATIONS ON THE REMOVAL OF GADOLINIUM FROM THE MODERATOR SYSTEM OF PRESSURISED HEAVY WATER REACTOR (PHWR) AND ADVANCED HEAVY WATER REACTOR (AHWR) V. Praveena*, Padma S. Kumar, A.L. Rufus and S. Velmurugan# Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities Kalpakkam (TN) – 603 102, INDIA *Department of Chemistry, University of Calicut, Calicut, INDIA # Corresponding Author: [email protected] Abstract Investigation on ion exchange removal of gadolinium taken as gadolinium nitrate, which is used as neutron poison in the moderator system of Pressurised Heavy Water Reactor (PHWR) and proposed to be used in Advanced Heavy Water Reactor (AHWR) was carried out. Mixed bed operation consisting of (a) strong acid cation resin (SAC) and strong base anion resin (SBA) and (b) strong acid action resin and acrylic acid based nitrate loaded weak base anion resin were employed for the removal gadolinium from its aqueous solution at pH 5. In the former case, the outlet of the mixed bed was highly alkaline, which resulted in precipitation of gadolinium hydroxide. In the latter case, the pH of the system never crossed 6 and gadolinium was effectively picked up on the resin without getting precipitated. Series operation consisting for strong acid cation resin followed by mixed bed column consisting of strong acid cation resin and strong base anion resin/acrylic acid based weak base anion resin was also investigated. In the first case where strong base anion resin was used, there was precipitation in the system owing to the increase in pH while in the case where weak base anion resin was used there was no problem of precipitation and gadolinium removed effectively and the pH was around 6. 75 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-180 CHEMISTRY MANAGEMENT OF GENERATOR STATOR WATER SYSTEM N. Sankar, V.S. Santhanam, S.R. Ayyar, P. Umapathi, P. Jeena, K. Hari Krishna, D.Rajendran Madras Atomic Power Station, Kalpakkam # Corresponding Author: [email protected] Abstract Chemistry management of water cooled turbine generators with hollow copper conductors is very essential to avoid possible re-deposition of released copper oxides on stator windings, which otherwise may cause flow restrictions by partial plugging of copper hollow conductors and impair cooling. The phenomenon which is of more concern is not strictly of corrosion failure, but the consequences caused by the re-deposition of copper oxides that were formed by reaction of copper with oxygen. There were also some Operating experiences (OE) related to Copper oxide fouling in the system resulting shut down/off-line of plants. In Madras Atomic Power Station (MAPS), the turbine generator stator windings are of Copper material and cooled by demineralized water passing through the hollow conductors. The heated water from the stator is cooled by process water. A part of the stator water is continuously passed through a mixed bed polisher to remove any soluble ionic contaminants to maintain the purity of system water and also maintain copper content as low as possible to avoid possible re-deposition of released copper oxides on stator windings. The chemistry regime employed is neutral water with dissolved oxygen content between 1000-2000ppb. Chemistry management of Stator water system was reviewed to know its effectiveness. Detailed chemical analyses of the spent resins from the polishing unit were carried out in various campaigns which indicated only part exhaustion of the polishing unit resins and reasonably low levels of copper entrapment in the resins, thus highlighting the effectiveness of the in-practice chemistry regime. 76 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-181 STUDIES WITH SOLID CHLORINE CHEMICAL FOR CHLORINATION OF SEA WATER SYSTEMS N.Sankar, P. Kumaraswamy, V.S. Santhanam, P. Jeena, K. Hari Krishna#, D. Rajendran, Madras Atomic Power Station, Kalpakkam # Corresponding Author: [email protected] Abstract Chlorination is one of the conventional methods to control biofouling of condenser cooling water systems using either river water, reservoir water or sea water. However, there are many safety concerns associated with handling, storage and application of gaseous chlorine. Studies were carried out with suitable alternativee chlorine chemical compounds which do not involve majority of these concerns but meet the functional requirement of gas chlorine. Trichloroisocyanuric Acid (TCCA) is one of the suitable alternatives to Gas chlorine. TCCA is a chlorine stabilized compound, stabilized with Cyanuric acid, thus similar to Gas Chlorine in its functions except that it is available in solid form. Release of chlorine is a gradual process in TCCA unlike Gaseous chlorine. Field studies with TCCA indicated gradual and near uniform release rate of chlorine, for longer duration with the requisite free residual chlorine levels (FRC). Thus, use of TCCA could be considered as a suitable alternative for gas chlorine for regular chlorination requirements. 77 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-182 CORROSION RATE OF CARBON STEEL IN NEUTRON SHIELD TANK WATER R. Ramakrishnan#, N. Rathinasamy and K. V. Ravi PRPD, Kalpakkam # Corresponding Author: [email protected] Abstract Neutron shield tank (NST) is an open tank of 12.5 meters height and 12 meters dia constructed around the reactor and filled water to provide sufficient shielding from the neutron radiation, to absorb the heat from the Containment Pressure suppression system during LOCA and to act as heat sink. NST is made of IS2062 carbon steel and it contains the stainless steel tanks, CS support structures, forged carbon steel gas cylinders, steel containment and its supports and emergency cooling down system condensers made of ASTM 350 gade LF2 carbon steel .All the equipments/systems located inside NST are painted with epoxy paint. NST is filled up 12meters ie with1200 m3 of water. The water chemistry parameters and microbiological parameters and corrosion rate of carbon steel materials in NST water at various water chemistry and various depths are discussed in the paper 78 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-183 OPTIMUM THICKNESS EVALUATION OF ZRO2 COATING ON TYPE 304L STAINLESS STEEL FOR CORROSION PROTECTION Nidhi Garg, Santanu Bera, V. S. Tripathia, Vijay Karkib and S. Velmurugan Water and Steam Chemistry Division, Bhabha Atomic Research Centre Facilities, Kalpakkam a Radiation and Photochemistry Division, BARC, Mumbai b Fuel Chemistry Division, BARC, Mumbai . # Corresponding author: [email protected] Abstract Nano-crystalline ZrO2 coatings of different thicknesses have been grown on preoxidized stainless steel (SS) surface by hydrothermal method in an autoclave. Thickness of the coating has been enhanced by repeating the deposition process several times using same precursor concentration. Several cycles of the deposition process lead to the increase of the coating thickness from 200 nm to ~1 m after the fourth cycle. The samples after different cycles of the coating have been extensively characterized by SEM-EDS technique to find the surface topography, coating thickness and composition. Corrosion resistance properties of the plain SS, pre-oxidized SS and all the ZrO2 coated samples were studied by potentiodynamic polarization technique and electrochemical impedance spectroscopy (EIS). Corrosion current densities (Icorr/cm2) of the coated samples are found to reduce significantly with the increase in thickness. After a certain critical thickness, the corrosion current density observed to attain a stable value. The coating was found to be continuous but porous after the first cycle but porosity of zirconia coating have been reduced drastically after the second cycle itself. EIS analysis confirms that the zirconia coated samples show insulating, barrier like characteristics in terms of high charge transfer resistance after the second cycle of zirconia deposition. The role of pre-oxidized surface composition and the interface between the pre-oxidized surface and the coating has been discussed in details by showing the depth distribution of Zr in the coating. 79 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-184 IODINE REMOVAL IN CONTAINMENT FILTERED VENTING SYSTEM DURING NUCLEAR ACCIDENT SubrataBera#, D. B. Nagrale, Anuj Kumar Deo, U. K. Paul, M. Prasad, A. J. Gaikwad Nuclear Safety Analysis Division, Atomic Energy Regulatory Board Niyamak Bhavan, Anushaktinagar, Mumbai-400094 # Corresponding author: [email protected] Abstract Post Fukushima nuclear accident, containment filtered venting system (CFVS) is being introduced in Indian NPP to strengthen the defense in depth safety barrier by reducing the containment pressure and ensuring the containment of potential radio-nuclides released during a severe accident. Radioactive iodine ( e.g. I-131, I-132, I-133, I-134, I-135 , etc.) is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the proposed design of CFVS, wet scrubbing mechanism has been employed where in iodine will be removed through chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper various regulatory aspects of CFVS system have been brought out such as validation of CFVS system, scrubber efficiency, measurement techniques involved, radiological impact assessment, radiation shielding requirement, etc. Analysis related to verification of the CFVS has been carried out through the estimation of full core inventory of iodine along with its isotopic distribution for a typical boiling water reactor. Keywords: CFVS, wet scrubber, Radioactive Iodine 80 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-186 AN OPERATIONAL EXPERIENCE WITH COOLING TOWER WATER SYSTEM IN CHILLING PLANT Manju B Rajan#, Ankan Roy, KV Ravi PRPD, BARC, Kalpakkam – 603102 # Corresponding author: [email protected] Abstract Cooling towers are popular in industries as a very effective evaporative cooling technology for air conditioning. Supply of chilled water to air conditioning equipments of various plant buildings and cooling tower water to important equipments for heat removal is the purpose of chilling plant at PRPD. The cooling medium used is raw water available at site. Water chemistry is maintained by make-up and blowdown. In this paper, various observations made during plant operation and equipment maintenance are discussed. The issues observed was scaling and algal growth affecting the heat transfer and availability of the equipment. Corrosion related issues were observed to be less significant. Scaling indices were calculated to predict the behavior. 81 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-187 CONTAINMENT BEHAVIOR DURING MOLTEN CORIUM CONCRETE INTERACTION Anuj Kumar Deo#, S. P. Lakshmanan, S. Bera, Balbir K. Singh, P. K. Baburajan, R. S. Rao, U. K. Paul & A. J. Gaikwad Nuclear Safety Analysis Division, Atomic Energy Regulatory Board Niyamak Bhavan, Anushaktinagar, Mumbai-400094 # Corresponding author: [email protected] Abstract During a severe accident in a NPP involving core melt and vessel-failure, the molten corium may fall in the cavity where it can react with the concrete basemat leading to the phenomena known as Molten Corium Concrete Interaction (MCCI). Due to the decay heat of the fission products, the un-cooled hot corium can cause ablation and decomposition of the concrete and thus penetrate the wall of the basemat which may result in loss of containment integrity. This may offer a potential route for release of radioactive materials to the soil and into the environment. Concrete is mainly composed of: SiO2, CaCO3 and H2O. Decomposition of concrete during heat-up starts with evaporation of physically bound water around 100°C. Dehydration of chemically bound water occurs up to 550°C. Decarbonation of CaCO3 (CaCO3 =>CaO + CO2) from the cement and carbonate aggregates occurs approximately between 700°C and 900°C. Liquid phases start to form between 1100°C and 1450°C. During concrete ablation, and corium oxidation, release of non-condensable gases (H2, CO, CO2) takes place into the containment and these gases may cause over pressurization of the containment. In the present work, a simplified analytical model of a typical BWR containment has been developed for ASTEC code to study the effect of MCCI on containment. The mass, physical composition and condition of the corium (debris introduced in ASTEC model) have been computed using the RELAP/SCDAP code. Rate of generation of hydrogen and noncondensable gases are obtained using ASTEC for postulated initiating event of LOCA with failure of ECCS. Keywords: MCCI, Containment, ASTEC, MEDICIS, Severe Accidents 82 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-188 DEUTERISATION OF MIXED BED ION EXCHANGE RESIN: KINETICS STUDY Satinath Ghosh1*, M. K. Tripathy1, Kajal Dhole1*, T. Vasudevan1, Satyam Shukla2 and R. S. Sharma1 Research Reactor Services Division1, Reactor Operations Division2, Bhabha Atomic Research Centre, Trombay, Mumbai-85. *Corresponding Author: [email protected] (Kajal Dhole); [email protected] (Satinath Ghosh) Abstract The process of deuterisation of a mixture of strongly acidic cationic and strongly basic anionic resins in a mixed bed system has been investigated for kinetics measurement through laboratory scale experiment. The up-flow fluidization method employing a heavy water flow from the bottom end of the mixed bed column at a reasonably low flow rate has been amply exploited for displacement of light water molecules inside the resin pores and adhering to resin surface as well. The course of deuterisation has been tracked down by determination of D2O content as a function of time and the process is found to exhibit a breakthrough type sigmoidal kinetics. An empirical relation, involving half-life of deuterisation and some process parameters such as flow rate, volume of light water to be replaced, could be achieved for plant scale deuterisation of a mixed bed ion exchanger prior to use in purification unit of heavy water process system of a nuclear reactor. 83 Symposium on water chemistry and corrosion in nuclear power plants in Asia-2015 AWC-189 FEASIBILITY STUDY ON NANO-STRUCTURED COATINGS TO MIGATE FLOWACCELERATED CORROSION OF CARBON STEEL PIPING SYSTEM Seunghyun Kim, Jeong Won Kim and Ji Hyun Kim# Ulsan National Institute of Science and Technology UNIST-gil 50, Eonyang-eup, Ulju-gun, Ulsan, Republic of Korea # email Corresponding Author: [email protected] Abstract To mitigate or prevent the flow-accelerated corrosion (FAC) of carbon steel pipes in secondary system of nuclear power plants, nano-structured coatings were adopted to supressed the dissolution of ferrous and magnetite ions. As candidates, TiO2 nano-particle reinforced electroless nickel plating and high-velocity oxy-fuel (HVOF) sprayed Fe-based amorhpous metallic coating (AMC) were selected and in order to evaluate their microstructures, electrochemical properties and FAC resistance characteristics using electron micrscopes, potentiodynamic polarization, electrochemical impedance spectrscopy and rotating cylinder autoclave systems. Microstructure analysis showed that TiO2 nano-structures and nano-crystalline structures were observed in Ni-P-TiO2 and HVOF Fe-based AMC, respectively. These structures induces the improved electrochemical properties according to high-temperature electrochemical experiments because diffusion of aggressive impurities were suprressed. As further work, the FAC simulation experiments using rotating cylinder autoclave system will be carried out. 84 Name Country email Organization Akhilesh C Joshi India [email protected], Bhabha Atomic Research Centre Alphonsa Joseph India Amit Ravindra K India [email protected], [email protected] [email protected] Amitava Roy India [email protected] Bhabha Atomic Research Centre Ananthan P India [email protected] Bhabha Atomic Research Centre Anil Pathrose India [email protected] Bhabha Atomic Research Centre Ankan Roy India [email protected] Bhabha Atomic Research Centre Anuj Kumar Deo India [email protected] Atomic Energy Regulatory Board Anupkumar B Ashwani Maheshwari Babu S India India India [email protected] - Bhabha Atomic Research Centre Nuclear Power Corporation of India Limited KARP, Kalpakkam Bairwa K K India [email protected] Bhabha Atomic Research Centre Balaji Gupta India [email protected] L&T , India Balaji V India [email protected] Bhabha Atomic Research Centre Barath Kumar S India [email protected] Indira Gandhi Centre for Atomic Research Basuki Baral India Biplob paul India [email protected] CWMF, Kalpakkam Brij Mohan India Nuclear Power Corporation of India Limited Chandramohan P India Chandran T J India [email protected] [email protected], [email protected] [email protected] Chellapandi P India [email protected] Bharatiya Nabhikiya Vidyut Nigam Ltd Cho Jae Seon Korea [email protected] FNC Tech Christoph Stiepani Germany [email protected] AREVA GmbH, Dash S C India [email protected] Nuclear Power Corporation of India Limited Debasis Mal India [email protected] Bhabha Atomic Research Centre Deepa Papachan India [email protected] Nuclear Power Corporation of India Limited Dey G R India [email protected] Bhabha Atomic Research Centre Dharuman S India [email protected] Dipankar Nanda India [email protected] Dineshkumar India [email protected]. Bhabha Atomic Research Centre Raja Ramanna Centre for Advanced Technology BioLogic Science Instruments Pvt. Ltd. Dong Seok Lim Korea [email protected] FNC Tech., Das N P I India [email protected] Indira Gandhi Centre for Atomic Research Francis Vincent India [email protected] Bhabha Atomic Research Centre Ganapathysubramanian India [email protected] Indira Gandhi Centre for Atomic Research Ganesh S India [email protected] Nuclear Power Corporation of India Limited George P J India [email protected] Bhabha Atomic Research Centre George R.P. India [email protected] Indira Gandhi Centre for Atomic Research Gopal Grandhi India [email protected] Atomic Energy Regulatory Board Hari Krishna K India [email protected] Nuclear Power Corporation of India Limited Harinath Y V India [email protected] Bhabha Atomic Research Centre Hazra G India [email protected] Burdwan University Hee-Sang Shim Korea [email protected] Korea Atomic Energy Research Institute Hee-Sang Shim Korea [email protected] Korea Atomic Energy Research Institute Helmut Nopper Germany [email protected] AREVA GmbH Hiren Joshi M India [email protected] Bhabha Atomic Research Centre Hur D.H. Korea [email protected] Korea Atomic Energy Research Institute Jagatap B N India [email protected] Bhabha Atomic Research Centre [email protected] Institute for Plasma Research Bhabha Atomic Research Centre Atomic Energy Regulatory Board Bhabha Atomic Research Centre Bhabha Atomic Research Centre Name country email Organization Jayasree Sriram India [email protected] Bhabha Atomic Research Centre Jaymin Gandhi India [email protected] Adani Infra India Ltd Jeena P India [email protected] Nuclear Power Corporation of India Limited Kaikondan A India [email protected] Bhabha Atomic Research Centre Kamachi mudali U India [email protected] Indira Gandhi Centre for Atomic Research Kazushige Ishida Japan [email protected] Hitachi Ltd. Kenji Hisamune Japan [email protected] The Japan Atomic Power Company Keny S J India [email protected] Bhabha Atomic Research Centre Kiran Kumar Reddy G India [email protected] Bhabha Atomic Research Centre Koteeswaran T J India [email protected] Nuclear Power Corporation of India Limited Krishna Mohan T V India tvkm@@igcar.gov.in Bhabha Atomic Research Centre Lakshmanan S P India [email protected] Atomic Energy Regulatory Board Lalu Thomas India [email protected] L&T , India Laxmi Narayan Gupta India [email protected] Institute for Plasma research, Madasamy P India [email protected], Bhabha Atomic Research Centre Mahendra Prasad India [email protected] Atomic Energy Regulatory Board Malathy Nagarajan India [email protected] Bhabha Atomic Research Centre Manjanna J India [email protected] Rani Channamma univeristy Manju B Rajan India [email protected] Bhabha Atomic Research Centre Manju Gupta India [email protected] AREVA GmbH Maruthu Pandiya Raja S India [email protected] Bhabha Atomic Research Centre Mathur P K India [email protected] Retd. Bhabha Atomic Research Centre Meng-Jen Chen Taiwan [email protected] Taiwan Power Company Mishra A. K India [email protected] CWMF, Kalpakkam Mishra H India [email protected] Bhabha Atomic Research Centre Muktibodh U C India Mohanty A K India [email protected] Indira Gandhi Centre for Atomic Research Mukunthan M India [email protected] Bhabha Atomic Research Centre Murugan M R India [email protected] Bhabha Atomic Research Centre Muthukumaran N India [email protected] NDDP, Kalpakkam N. Jayaraman India [email protected] Bhabha Atomic Research Centre Narasimhan S V India [email protected] Retd. Bhabha Atomic Research Centre Natarajan E India [email protected] Anna University Nidhi Garg, India [email protected] Bhabha Atomic Research Centre Nilesh Patel India [email protected] Adani Infra India Ltd Osamu Shibasaki Japan [email protected] Toshiba Corporation Padala Abdul Nishad India [email protected] Bhabha Atomic Research Centre Padhi R K India [email protected] Indira Gandhi Centre for Atomic Research Padma S.Kumar India [email protected] Bhabha Atomic Research Centre Pal P K India [email protected] Nuclear Power Corporation of India Limited Panigrahi B S India [email protected] Indira Gandhi Centre for Atomic Research Pankaj Wani India Prak Byeong-Ho Korea [email protected] Prabhakar Jain India Prasad Y V D India Pushpalata Rajesh India [email protected] Prasadgowda.mechanical.98@gmail. com [email protected] Nuclear Power Corporation of India Limited KEPCO Engineering & Construction company Heavy Water Board Rachna N. Dave India [email protected] Nuclear Power Corporation of India Limited Nagarjuna university Bhabha Atomic Research Centre Bhabha Atomic Research Centre Name country email Organization Radhakrishnan R K India [email protected] Nuclear Power Corporation of India Limited Raghavendra India [email protected] Bhabha Atomic Research Centre Rajamohan R India [email protected] Bhabha Atomic Research Centre Rajesh Kumar India [email protected] Bhabha Atomic Research Centre Rajput M M India [email protected] Bhabha Atomic Research Centre Rajnish Kumar India Ramakrishnan R India [email protected] Bhabha Atomic Research Centre Rangarajan S India [email protected] Bhabha Atomic Research Centre Ranjana Kusari India Rao K V India [email protected] Bhabha Atomic Research Centre Ravi K V India [email protected] Bhabha Atomic Research Centre Ravidranath India [email protected] Nuclear Power Corporation of India Limited Rout D India [email protected] Nuclear Power Corporation of India Limited Rufus A L India [email protected] Bhabha Atomic Research Centre Sahu B S India [email protected] Nuclear Power Corporation of India Limited Santanu Bera India [email protected] Bhabha Atomic Research Centre Santhakumar M India [email protected] Bhabha Atomic Research Centre Saravanan T India [email protected] Bhabha Atomic Research Centre Sathyaseelan V S India [email protected] Bhabha Atomic Research Centre Satinath Ghosh India [email protected] Atomic Energy Regulatory Board Selvam T India [email protected] Bhabha Atomic Research Centre Selvavinayagam P India [email protected] Nuclear Power Corporation of India Limited Sengupta B India [email protected] Nuclear Power Corporation of India Limited Seshadri H India [email protected] Seunghyun Kim Korea [email protected] Sharma R S India [email protected] Atomic Energy Regulatory Board Ulsan National Institute of Science and Technology, Bhabha Atomic Research Centre Shetty P S India [email protected] Bhabha Atomic Research Centre Shiv Raj Saran India [email protected] Bhabha Atomic Research Centre Shivakamy K India [email protected] CWMF, Kalpakkam Shreekumar B India [email protected] KARP, Kalpakkam Shruti Aich India [email protected] Bhabha Atomic Research Centre Sinu Chandran India [email protected] Bhabha Atomic Research Centre Sreevidya N India [email protected] Indira Gandhi Centre for Atomic Research Srinivasa Rao G India - KARP, Kalpakkam Srinivasan G India [email protected] Indira Gandhi Centre for Atomic Research Srinivasan M P India [email protected] Bhabha Atomic Research Centre Sriyuthamurthy P India [email protected] Bhabha Atomic Research Centre Subba Rao T India [email protected] Bhabha Atomic Research Centre Subrajit Tiwari India Subramanian H India [email protected] Bhabha Atomic Research Centre Subratabera India [email protected] Atomic Energy Regulatory Board Sudakar Rao A India [email protected] Heavy Water Board Sudha R India [email protected] Indira Gandhi Centre for Atomic Research Sudhir K. Shukla India [email protected] Bhabha Atomic Research Centre Sujit Basak India Sumathi Suresh India [email protected] Bhabha Atomic Research Centre Suresh K India [email protected] Nuclear Power Corporation of India Limited Suriyanarayanan A India [email protected] Indira Gandhi Centre for Atomic Research Atomic Energy Regulatory Board Nuclear Power Corporation of India Limited AUGF, Kalpakkam Bhabha Atomic Research Centre Name country email Organization Surya Rao S India [email protected] Heavy Water Board Swapan Kumar Mahato India [email protected] Indira Gandhi Centre for Atomic Research Thorat D D India [email protected] Bhabha Atomic Research Centre Tripathi V S India [email protected] Bhabha Atomic Research Centre Upadhyay S K India [email protected] Nuclear Power Corporation of India Limited Veena Subramanian India [email protected] Bhabha Atomic Research Centre Velmurugan S India [email protected] Bhabha Atomic Research Centre Venkatesh P India [email protected] NDDP, Kalpakkam Venkatraman B India [email protected] Indira Gandhi Centre for Atomic Research Venugopalan V P India [email protected] Bhabha Atomic Research Centre Venkata Rao Naidu India [email protected] Nuclear Power Corporation of India Limited Venkatesh M India [email protected] BioLogic Science Instruments Pvt. Ltd. Vijayalakshmi S India [email protected] Indira Gandhi Centre for Atomic Research Vikrant Gupta India [email protected] Institute for Plasma Research Vinay Chaturvedi India [email protected] KNRPC, Kalpakkam Vinita Vishwakarmaa India [email protected] Sathyabama University Vivekanand Dubey India [email protected] Bhabha Atomic Research Centre Viveknand Kain India [email protected] Xinqiang Wu China [email protected] Yadav R N India [email protected] Bhabha Atomic Research Centre Institute of Metal Research, Chinese Academy of Sciences KNRPC, Kalpakkam Yasuhiro Chimi Japan [email protected] Japan Atomic Energy Agency Yaw-Ming Chen Taiwan [email protected] Industrial Technology Research Institute, Yosuke Katsumura Japan [email protected], [email protected] Japan Radioisotope Association Yusa Muroya Japan [email protected] Yutaka Watanabe Japan [email protected] Institute of Scientific and Industrial Research, Osaka University, Tohoku University