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Fast Reactor Physics and Technology

Report of the Technical Meeting

International Atomic Energy Agency, Vienna, Austria

hosted by

IGCAR Kalpakkam, India

14 – 18 November 2011

Reproduced by the IAEA Vienna, Austria, 2011

NOTE The material reproduced here has been supplied by the authors and has not been edited by the IAEA. The views expressed remain the responsibility of the named authors and do not necessarily reflect those of the government(s) of the designating Member State(s). In particular, neither the IAEA nor any other organization or body sponsoring the meeting can be held responsible for this material

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CONTENTS

1. FOREWORD .................................................................................................................... 5

2. EXECUTIVE SUMMARY .............................................................................................. 7

3. OBJECTIVES OF THE MEETING ................................................................................. 8

4. SUMMARY OF SESSIONS ............................................................................................ 9

4.1 Status of national programs ................................................................................ 9 4.1.1 Belgium (presented by Didier De Bruyn) ............................................... 9

4.1.2 China (presented by Keyuan Zhou) ........................................................ 9

4.1.3 Czech Republic (presented by Milan Brumovsky) ............................... 10 4.1.4 European Commission (presented by Kamil Tucek) ............................ 10 4.1.5 France (presented by Francois Gauché) ................................................ 11

4.1.6 Germany (presented by Bruno Merk) ................................................... 11

4.1.7 India (presented by P. Chellapandi) ...................................................... 11

4.1.8 Italy (presented by Alessandro Gessi) .................................................. 12

4.1.9 Japan (presented by T. Asayama) ......................................................... 12

4.1.10 Republic of Korea (presented by Sun Rock Choi) ............................... 12

4.1.11 The Netherlands (presented by Ferry Roelofs) ..................................... 12

4.1.12 Russian Federation (presented by A.P. Sorokin) .................................. 13

4.1.13 Sweden (presented by Janne Wallenius) .............................................. 13

4.1.14 Ukraine (presented by Sergei Fomin) ................................................... 13

4.2 Reactor Physics and Core Design ..................................................................... 14

4.3 Advanced Reactor Designs including ADS ...................................................... 14

4.4 Structural Materials Development .................................................................... 15

4.5 Coolant Technology and Component Development ........................................ 16

4.6 Thermal Hydraulics .......................................................................................... 18 4.7 Reactor Safety ................................................................................................... 19

5. SUMMARY OF PRESENTATIONS ............................................................................. 20

5.1 S. Fomin (KIPT, Kharkov, Ukraine) ................................................................ 20

5.2 D. Sunil Kumar (IGCAR, Kalpakkam, India) .................................................. 20

5.3 B. Merk (HZDR, Rozendorf, Germany) ........................................................... 21

5.4 P. Wolniewicz (Uppsala University, Uppsala, Sweden) .................................. 21

5.5 F. Klassen (NRG, Petten, The Netherlands) ..................................................... 22

5.6 T. Satyasheela (IGCAR, Kalpakkam, India) .................................................... 23

5.7 D.K. Mohapatra (IGCAR, Kalpakkam, India) ................................................. 23

5.8 G.Raghu Kumar (IGCAR, Kalpakkam, India) ................................................. 24

5.9 P.Le Coz (CEA, France) ................................................................................... 25 5.10 D. de Bruyn (SCK•CEN, Mol, Belgium) ......................................................... 25

5.12 P. Puthiyavinayagam (IGCAR, India) .............................................................. 26

5.13 V. Rajan Babu (IGCAR, India) ........................................................................ 27

5.14 R.Vijayashree (IGCAR, Kalpakkam, India) ..................................................... 27

5.15 S. Raghupathty (IGCAR, Kalpakkam, India) ................................................... 28

5.16 V. Balasubramaniyan (IGCAR, Kalpakkam, India) ......................................... 28

5.17 K. Madhusoodhanan (IGCAR, Kalpakkam, India) .......................................... 29

5.18 S.B. Degweker (BARC, Mumbai, India) .......................................................... 29

5.20 T. Jayakumar (IGCAR, Kalpakkam, India) ...................................................... 30

5.21 B.K. Panigrahi (IGCAR, India) ........................................................................ 31

5.22 V. Kathik (IGCAR, India) ................................................................................ 32

5.23 T. Asayama (JAEA, Ibaraki, Japan) ................................................................. 33

5.24 C. Latge (CEA, France) .................................................................................... 34 5.25 A. Gessi (ENEA, Italy) ..................................................................................... 34 5.27 R. Ganesan (IGCAR, India) ............................................................................. 34

5.28 N. Devictor (CEA, France) ............................................................................... 35

5.29 K.K. Rajan (IGCAR, India) .............................................................................. 36

5.30 P. Selvaraj (IGCAR, India) ............................................................................... 36 5.31 R.Sridharan aj (IGCAR, India) ......................................................................... 37

5.32 P.Selvaraj (IGCAR, India) ................................................................................ 38 5.33 R. Sridharan (IGCAR, India) ............................................................................ 38

5.34 V.A. Suresh Kumar (IGCAR, India) ................................................................ 40

5.35 B.K. Sreedhar (IGCAR, India) ......................................................................... 40

5.36 V. Prakash (IGCAR, India) .............................................................................. 41

5.37 F. Roelofs (NRG, The Netherlands) ................................................................. 41

5.38 R.Vijayashree (IGCAR, India) ......................................................................... 42

5.39 S.R. Choi (KAERI, Republic of Korea) ........................................................... 43

5.40 S. Bortot (Politecnico di Milano, Italy) ............................................................ 43

5.41 K.Velusamy (IGCAR, India) ............................................................................ 44

5.42 A.P.Sorokin (IPPE, Russia) .............................................................................. 45

5.43 K. Tucek (IET, JRC, European Commission) .................................................. 46

5.44 K.Velusamy (IGCAR, India) ............................................................................ 47

5.45 V.Vinod (IGCAR, India) .................................................................................. 48

5.47 E. Hemanth Rao (IGCAR, India) ..................................................................... 48

5.48 S. Kumar Das (IGCAR, India) ......................................................................... 49

6. RECOMMENDATIONS FOR REAL TIME RESEARCH AREAS ............................. 50

6.1 Research Areas in Fast reactor physics and technology ................................... 50

7. ANNEXES ...................................................................................................................... 51

7.1 Agenda of the meeting ...................................................................................... 51 7.2 List of participants ............................................................................................ 56

1. FOREWORD

Renewed interest in nuclear energy is driven by the need to develop carbon free energy sources, demographics and development in emerging economies, and security of supply concerns, is likely to accelerate fast reactor Research and Development (R&D) activities and deployment of new prototypes. Today, several IAEA Member States are active in the area of fast reactor technology development, and it appears realistic to assume that deployment of advanced prototype fast power reactors will materialize within the time frame of the next 10 - 20 years. It is expected that an initial phase of demonstration will be followed by a transition phase to commercial fast reactors, and eventually to a further innovation step opening up the full potential of the fast neutron system and closed fuel cycle technologies. Future deployment of such innovative systems will enhance sustainability and reduce radioactive wastes of the nuclear fuel cycle, further enhance reactor safety, improve economics for electricity production and new applications such as the supply of process heat and/or hydrogen production, and increase proliferation resistance.

The IAEA support of fast reactors R&D and deployment include Coordinated Research Projects (CRPs), Technical Working Group (TWG-FR) meetings, topical meetings, experts reviews and information dissemination, in the form of the fast reactors database as well as IAEA publications. These activities help to build an effective platform for information exchange at an international level in order to summarize available data on operational parameters, physical, thermo-hydraulic and thermo-mechanical characteristics, technological requirements, methods and criteria to ensure safe operation; design data like dimensions, materials information, and main design features and performance parameters of fast reactor cores, components, and various systems, along with sketches and drawings.

As a consequence of the Fukushima accident, further incorporation of inherent and passive safety design features are essential in order to reduce design complexity of new reactors, as well as the need for human intervention resulting in fewer potentially dangerous actions. However, taking into consideration recent advances in design of innovative fast reactor technology with passive safety features, improved fuel cycle and high performance, further sharing of experiences and information on reactor design and its critical components, as well as various technical parameters, boundary conditions, and data related to reactor core and structural materials is needed. Specifically, recent advances in research and technology development in the area of reactor core characteristics, fuel design, and performance of sodium, lead, and gas cooled fast reactor designs will be discussed during the Technical Meeting. In particular emphasis will be placed on design aspects of advanced prototype fast reactors, in which much of the scaling up required for a commercial station in terms of both overall size and individual components, as well as the testing of innovative features needs to be done.

The operating regimes for structural components materials in these innovative reactor concepts extend to higher temperatures and doses than the experience in existing power reactors. In order to achieve these objectives, a concerted effort is still needed to develop and qualify new materials for key structural components. At present, extensive characterization and testing of recently developed candidate materials using the most advanced techniques of as-fabricated materials are needed as a necessary precursor to the subsequent understanding of the radiation degradation, ageing and phase stability mechanisms. In addition, improved design methodologies to ensure the safe application of these advanced materials in high temperature reactor environments is needed. This will also require the development of

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advanced monitoring and inspection provisions of inner and outer surface of the primary vessel and internal structures and other primary components.

In total, 67 international experts from Belgium, China, Czech Republic, European Commission, France, Germany, India, Italy, Japan, Republic of Korea, Netherlands, Russia, Sweden, and Ukraine were present. The meeting was chaired by Mr S.C. Chetal and Mr. P. Chellapandi was appointed IGCAR organiser in charge of the meeting. The IAEA scientific officers responsible for this meeting were Mr A. Zeman of the Physics Section, Department of Nuclear Sciences and Applications and Mr S. Monti of the Nuclear Power Technology Development Section, Department of Nuclear Energy. The session chairs were S. Monti (IAEA), K.Tucek (IET, JRC-EC), A. Zeman (IAEA), P. Chellapandi (IGCAR), C. Latge (CEA), B. Merk (FZD Rosendorf), F. Gauche (CEA).

2. EXECUTIVE SUMMARY

This meeting focused on research and development, deployment, operation and safety aspects of various fast reactors designs. The IAEA Technical Meeting on Fast Reactor Physics and Technology provided a very good platform for exchange and sharing of the knowledge and experiences in various aspects of fast reactors. The experts agreed that the IAEA should continue in support activities in this specific but very important subject. It was clearly recognised that such a meeting with broader participation of experts from different communities help to strengthen and coordinate the international activities as well as stimulate new initiatives which are needed in the area of fast reactors. Experts agreed that appropriate attention has to be given to the further networking in order to achieve the consensus on design criteria of new fast reactor systems. This is an emerging task since several Member States have launched deployment and construction of commercial nuclear power plants based on fast reactor technology. In view of this the meeting identified specific subjects which are very important for future development of fast reactors, in particular further adaptation of safety criteria which are needed to cover fast reactors from a commercial deployment point of view. In addition, education and training is also considered to be an important topic to be supported by the Agency on a long term basis. It is very important to strengthen support specifically focused on a “new generation” of engineers and young professionals in order to preserve the .unique knowledge and technical information as this technology is foreseen to be a key technology in the near future.

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3. OBJECTIVES OF THE MEETING

The overall objective of the technical meeting, “Fast reactor physics and technology” was to present the state-of-the-art as far as innovation for reactor designs, in particular as far as critical components is concerned. The specific objectives of the meeting were to focus on a technological base, concepts and operation of such techniques with consideration of proposals for further scientific development in this area. Besides the overall objective, further specific issues addressed during the meeting included, in particular: - Summary of recent advances in the development of reactor designs for minor actinides

utilisation with emphasis on sodium and lead-cooled concepts. - Discuss the impact of novel core structural materials for reactor internals and fuel

assemblies from the neutronics and thermal-hydraulics viewpoints (reactivity, coolant circulation, heat capacity, etc.).

- Assess the innovative aspects of small and medium size sodium and heavy liquid metal cooled fast reactors, in particular long-life designs (no on-site refuelling), inherent safety characteristics, and “mobile fuels”.

- R&D activities related to sodium, heavy-liquid metal and gas cooled reactor concepts in terms of experiments, methods development as well as validation and qualification.

Following the experts’ presentations, working groups discussed the priorities in the areas of (1) transmutation of long lived actinides; (2) reassessment of core component materials (emphasis on metallic fuel), (3) long-life core designs; and (4) methods development experiments, as well as validation procedures.

4. SUMMARY OF SESSIONS

4.1 Status of national programs

There were 13 presentations by representatives of the participating Member States plus one by the representative of the European Commission, concerning the respective national programmes in the field of fast reactors.

4.1.1 Belgium (presented by Didier De Bruyn)

Belgium and in particular SCK.CEN activities related to fast spectrum systems are concentrated on the design of the MYRRHA facility, a 100 MWth lead-bismuth cooled irradiation facility supposed to operate in both critical and subcritical mode.

In 2010 the Belgian Federal Government decided to support the realisation of the MYRRHA project allocating 60 M€ for the period 2010 – 2014, in order to:

- Conduct R&D to minimize technological risks;

- Finalize front end engineering design;

- Secure licensing (preparing legal requirements);

- Implement a management structure at international level.

In 2014 instructions should be provided for proceeding to the next step and, if the green light is given, . Belgium will oversee .40% of the total investment costs.

MYRRHA is also supported by ESFRI (European Strategy Forum on Research Infrastructures), ESNII (European Sustainable Nuclear Industrial Initiative) and a number of projects carried out under the various European Framework Programmes.

In support of the physics of MYRRHA, SCK.CEN is performing the GUINEVERE experiment, a zero-power ADS facility consisting of a DT accelerator coupled with the VENUS reactor at Mol.

4.1.2 China (presented by Keyuan Zhou)

After the Fukushima Dai-ich NPP event, the Chinese authorities of nuclear power launched in-depth nuclear safety checks, but also reiterated that nuclear energy, as a clean and reliable energy, will play an increasingly important role in Chinese energy development, and has enacted the nuclear energy development policy of China, "Based on the principle of 'safety comes first', efficiently develop nuclear energy". After nearly 25 years of development, China has 11 nuclear power units in operation and 20 nuclear power units under construction.

Fast reactor technology research of China -started in the 1960s, and was part of the national high-tech development plan in the 1980s. The China Experimental Fast Reactor (CEFR) is a 20 MWe SFR which is using UO2 fuel at present and will be using MOX in the future. The CEFR achieved criticality for the first time in July 2010 and then finished the physics startup

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testing, and completed the first grid connecting test in July 2011.The raising power tests of CEFR is underway now.

China has now launched the research project of the large-scale SFR NPP construction, and the FBR commercialization is expected by 2025-2030 with realization of the CDFR (Demonstration Plant of 800 ~ 1000 MWe) and successively of the CCFR (Commercial Plant of more than 1000 MWe).

4.1.3 Czech Republic (presented by Milan Brumovsky)

A FR oriented national programme was carried out in former Czechoslovakia between 1975 and 1990 mostly oriented to qualification of materials to be used in Na environment. 8 Na loops were operated as well as a large thermal-hydraulic loop. Former Czechoslovakia also supplied the steam generators for Russian BOR60 reactor which are still in operation.

At present, Czech Republic is participating in several European Projects devoted to GENIV FRs as well as in the European Energy Research Alliance (EERA) on nuclear structural materials.

Czech Republic is also a member of the trilateral consortium supporting the realization of the European GFR experimental reactor called ALLEGRO.

4.1.4 European Commission (presented by Kamil Tucek)

The European Union (EU) in 2007, launched. a policy program, Strategic Energy Technology Plan (SET-Plan), to achieve ambitious policy targets by 2020 and 2050. The SET-Plan recognizes nuclear energy fission as one of the six priority technologies to achieve these targets.

Together with the SET-Plan, Sustainable Nuclear Energy Technology Platform (SNETP) was launched in 2007 to create impetus for enhancing nuclear fission in sustainable energy mix in Europe. It comprises of more than 100 members from industry, research, academia, technical support organisations (TSOs), and NGOs. One of the most important outputs of SNETP is the Strategic Research Agenda (SRA), which provides a roadmap for Research, Development & Demonstration of sustainable nuclear fission technologies. One of the pillars of the SRA is the development of GENIV fast reactors. The three technologies for European GENIV FRs selected by the European Sustainable Nuclear Industrial Initiatives (ESNII) are SFR (reference technology), LFR and GFR (alternatives). The corresponding proto and demo plants are: ASTRID for the SFR reference technology, MYRRHA and then ALFRED for the LFR and ALLEGRO for the GFR.

The European Commission also favoured the creation of a Joint Research Program on Nuclear Materials between the MSs, in the frame of the above mentioned EERA initiative.

4.1.5 France (presented by Francois Gauché)

Two fast neutron systems are developed in parallel in France in order to comply with the technological goals established by the Generation IV international forum:

- SFR as reference technology;

- GFR as a longer term option;

The SFR program is focused on the development of the ASTRID industrial prototype, a 600 MWe GENIV SFR expected to be in operation in France at the beginning of 2023. The present design phase is oriented to the preliminary choice of technical options; a decision to start building is anticipated in 2017. The ASTRID project is coordinated by CEA and at the moment supported by AREVA, EdF, ALSTOM, COMEX Nucléaire in view of the creation of an international consortium. The French government has already granted 650 M€ up to 2017.

4.1.6 Germany (presented by Bruno Merk)

After the Fukushima accident the German Government decided upon the immediate and permanent shut-down of 8 NPPs while the rest of the fleet (11 NPPs) are planned to be shutdown between 2015 and 2022. Nevertheless, an official document on energy policy . released recently confirmes that German commitment on nuclear research mainly focuses on safety and long term waste disposal (e.g. partitioning & transmutation) as well as on knowledge preservation and international collaboration.

An overview of the R&D activities carried out by Helmholtz-Zentrum Dresden-Rossendorf in particular in the field of fast reactor (extension of DYN3D, liquid metal technology, cross section measurements) was also presented.

4.1.7 India (presented by P. Chellapandi)

At present, India has a nuclear capacity of some 4183 MWe and expects to have 20,000 MWe nuclear capacity on line by 2020 and 63,000 MWe by 2032. It also plans to become a world leader in nuclear technology in particular due to its expertise in fast reactors and thorium fuel cycle.

The very ambitious FBR program which started in the ‘80s with the first criticality of the 13.5 MWe FBTR (Fast Breeder Test Reactor) in 1985 forsees the commissioning of the MOX fuelled 500 MWe PFBR (Prototype Fast Breeder Reactor) in 2013, followed by the deployment of three units of the MOX fuelled 500 MWe CFBR from 2023 and finally of the commercial metal fuelled 1000 MWe FBR beyond 2025.

The PFBR construction proceeds according to schedule and it is supported by an extensive experimental and qualification programme including core safety (thermal-hydraulic, mechanics, etc.), sodium fire, CDA analysis, development of sensors for Na application, ISI&R, etc. Some R&D activities are also intended to support the next phase represented by CFBR. India is also actively participating in various IAEA activities.

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4.1.8 Italy (presented by Alessandro Gessi)

The Italian program in the field of fast spectrum system is very much oriented to the development of the GEN-IV Lead-cooled Fast Reactor through a domestic program supported by the Minister of Economic Development and a broad participation in the European programmes and projects devoted to GENIV FRs. In particular Ansaldo Nuclear is the leader in the European project devoted to the development of the European LFR DEMO (ALFRED) and ENEA is performing various analytical and experimental activities concerning the physics and technology of the LFR.

4.1.9 Japan (presented by T. Asayama)

As far as GEN-IV FRs, the Japanese programme is focused on the development of the loop-type JSFR reactor with the vision of first operation at the horizon of 2025 and full commercial deployment towards 2050.

As a consequence of the Fukushima accident, the JSFR conceptual design has been postponed to 2 years and an increased effort is being made on safety issues.

A Decision from the Japanese Government is also expected. for the restarting of JOYO and MONJU reactors.

4.1.10 Republic of Korea (presented by Sun Rock Choi)

In the frame of its program for closing the fuel cycle based on the use of SFR and pyro-processing of metal fuel, RoK is developing a 600 MWe pool-type U-Zr (then U-TRU-Zr) fuelled GEN-IV SFR called KALIMER-600.

A large R&D supporting program is underway and includes: experimental Na loop for testing the main components, metallic fuel development, new compact Na/CO2 heat exchanger, V&V of core and safety analysis codes, under-sodium sensors, cross section adjustment, etc.

It is worth mentioning that KAEC has approved a long-term SFR R&D plan which foresees the construction of the DEMO plant by 2028.

4.1.11 The Netherlands (presented by Ferry Roelofs)

In the Netherlands a 480 MWe PWR which passed the European stress test is in operation and two applications for a second NPP has been filed. NRG operates the EC owned 45 MW HFR research reactor since 1961.

A number of analytical and experimental FR-oriented R&D activities have been carryied out by NRG in particular in the field of advanced thermal-hydraulic (CFD and system codes), MA-bearing fuel irradiation (see also presentation by F. Klaassen), innovative materials.

4.1.12 Russian Federation (presented by A.P. Sorokin)

Since 30 years the Russian Federation is operating the present most powerful fast reactor BN600 in Beloayarskaya. On the same site the MOX fuelled BN-800 is under an advanced construction phase. BN-800 is considered to be the bridge to the new generation nuclear technology in Russia.

In February. 2010 the Government of the Russian Federation confirmed the Federal Target Program “Nuclear Energotechnologies New Generation for 2010-2015 and on a prospect till 2020” which includes:

- Development of BREST-300 Lead-cooled Fast Reactor;

- Development of SVBR-100 LBE-cooled Fast Reactor;

- Development of the design of a new generation Sodium-cooled Fast Reactor of 1200 MWe, i.e. BN-1200;

- Development of integrated systems of new generation codes for analysis and substantiation of safety of innovative FR and related fuel cycle.

The BN-1200 design is supported by a very large analytical and experimental program including T/H modelling and testing, transient and safety analysis, Na purification, heat exchangers and steam generators, sensors in Na, etc.

In parallel R&D activities are also carried out in support of BREST-300 and SVBR-100.

4.1.13 Sweden (presented by Janne Wallenius)

FR-oriented R&D activities in Sweden are carried out by KTH, Chalmers and Uppsala Universities under the auspices of the GENIUS programme funded by the Government, as well as in the frame of Euratom projects.

Most of the effort is focused on the development of GENIV LFR and in particular on innovative fuel development (e.g. U nitride and Pu fuel fabrication), materials research (e.g. corrosion tests) and safety. In this context the ELECTRA – European Lead Cooled Training Reactor is being proposed. ELECTRA is a 0.5 MW core with (Pu,Zr)N fuel 100% cooled by natural circulation of lead mainly oriented to E&T and liquid metal reactor dynamics research. Discussion is in progress in order to realize ELECTRA in Oskarsham as well as for involving international partners.

4.1.14 Ukraine (presented by Sergei Fomin)

FR-oriented R&D activities in Ukraine are carried out mainly by National Science Center “Kharkov Institute of Physics and Technology” (NSC KIPT, Kharkov) and Kiev Institute for Nuclear Research (KINR, Kiev) under the auspices of National Academy of Science of Ukraine within the Program "Scientific and technical support of nuclear power development and application of radiation technologies in the fields of economy" funded by the Government.

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A number of analytical and experimental FR-oriented R&D activities have been carried out by NSC KIPT in particular in the field of development of new radiation-structural and functional materials for the needs of the nuclear industry. There are several installations for the reactor irradiation damage simulation based on the charged particle accelerators.

An agreement with the US Argonne National Laboratory has been signed for construction of subcritical assembly (ADS) driving by electron accelerator in NSC KIPT.

Ukraine is also working on the physics of nuclear burning wave reactor concept (traveling wave reactor) originally proposed by Russian Academician L.P. Feoktistov in 1988 (see presentation by S.P. Fomin below).

4.2 Reactor Physics and Core Design

The Session on Reactor Physics and Core Design provided a multi-faceted picture with presentations ranging from overviews of different fast reactor related activities in individual organisations to technical presentations on specific topics of fast reactor fuel development and irradiations, feasibility and performance assessments for innovative shield design, reactor physics and safety aspects of fast reactors with metallic fuel, and development as well as validation of neutronic and safety analysis codes. Feasibility assessments were also presented on fundamental, innovative concepts for the development of the instrumentation for the detection of coolant void bubbles in the reactor cores and on the neutronic feasibility of the travelling wave reactor.

The presentations showed a high degree of involvement and interest of individual organisations to pursue and investigate potentially important innovations for the improvement of safety and performance of fast reactors. In general, the on-going efforts on the development and qualification of the state-of-the-art computer tools for fast reactor design and safety assessments contribute to improvement of the analysis capabilities. These activities could be further strengthened in the framework of appropriate international collaborations.

4.3 Advanced Reactor Designs including ADS

The first presentation focused on an overview of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project, SFR prototype proposed by France and endorsed by the Sustainable Nuclear Energy Technology Platform (SNETP) as well as by the European Sustainable Nuclear Industrial Initiative (ESNII). ASTRID is an industrial prototype, which integrate long lasting French and international experience on SFR with the new GEN-IV requirements in terms of safety, operability, waste management in a close fuel cycle and economics. This project also benefits from industrial partnerships at the moment represented by large French nuclear industries (AREVA-NP, EDF, ALSTOM, COMEX NUCLEAIRE) but to be further extended to other international organizations. The pre-conceptual design started in 2010 and will be concluded in 2012 with some 1000 technical documents – including a safety oriented report with advice by the French Safety Authority - which will take a decision on launching the conceptual design. Some options concerning core design, ECS, number of loops, core catcher technology etc. are still open and investigated with the aim to fix them by the end of 2012. Safety requirements and characteristics will take into account WENRA requirements as well as lessons learned from

Fukushima. As for transmutation (Am and eventually Np and Cm) capability, ASTRID will follow a progressive approach. Go/Non-go decision will be taken at the end of 2014: at that stage the basic design will be launched. First criticality is anticipated by end 2023.

A second talk briefly introduced MYRRHA (Multiporpose hYbrid Research Reactor for High-tech Applications) facility, a 100 MWth lead-bismuth cooled multipurpose facility supposed to operate in both critical and subcritical mode, which has been launched by SCK.CEN-Mol. The concept of flexible fast spectrum irradiation facility represents subcritical configuration composed of a 600 MeV – 4 mA LINAC, a Pb-Bi spallation target and a LBE-cooled subcritical core of 65 to 100 MWth. The nuclear island design, as well as the plant lay-out have been evolving from 2000 to 2010 with the support of various projects carried out under the Euratom Framework Programmes. The present design allows the achievement of a fast neutron flux (E> 0.75 MeV) of 1015 n/cm2s at 100 MWth. In 2009 MYRRHA benefitted from a full facility lay-out including accelerator, beam transport line and auxiliaries systems.

The presentations that followed in this session covered... lessons learned from the construction of the Indian PFBR, as well as the way ahead towards the Indian CFBR development in terms of new design approach and technical developments. A very comprehensive presentation on how the PFBR construction experience can help in easing construction of CFBR was shortly discussed as well. The presentation covered feedback from material suppliers, manufacturers, fabricators and machinists as well as from system integrators, erectors, industrial safety, quality control, commissioning teams and maintainers. This presentation also provided an opportunity to appreciate, through several pictures, the huge constructional work of PFBR: from reactor vault to lateral erection, from sodium tanks installation to safety vessel and core catcher erection, from main vessel erection to roof and thermal baffle erection up to grid plate, inner vessel, roof slab, IHX, steam generator, primary pump, DHE, etc. manufacturing and installation.

A new design approach for future Indian FBRs in the area of reactor core, reactor assembly, shutdown systems, fuel handling systems, heat transport systems, Instrumentation & control was briefly introduced by IGCAR experts. Most of the presentations elucidated the CFBR design features retained from PFBR as well as the innovative ones which allow optimisation, simplification, higher performances and cost and time reductions. The very last presentation of the session summarized Physics studies on thorium utilisation and subcritical measurements in Accelerator Driven Systems at BARC. The ADS development is embedded in the general Indian nuclear energy scenario which accounts for very limited U availability, small Pu inventory, present small volume of nuclear waste and large Th availability. ADS are regarded as nuclear systems which can achieve faster breeding, flexibility in use of fuels and simplification of Th fuel cycle. The next step is the development and realization of a subcritical facility at Purnima, intended to qualify various methods of subcriticality measurements as well as different factors affecting measurements.

4.4 Structural Materials Development

The session of Structural Materials development for fast rectors addressed various pending issues of SFR and GFR technology. Individual talks were focused on research and technological issues of structural materials and on-going R&D programs related to the fast-reactor reactor and primary components. That included the talk on GFR road map, where a short introduction of the European demonstrator ALLEGRO was given. The GFR

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demonstrator reactor ALLEGRO is being discussed in Central Europe, specifically Czech republic, Slovakia and Hungary. In terms of the priorities it has priority no. 3 and currently it is under pre-conceptual design. According to the most recent roadmap the construction is scheduled to be initiated by 2018. In view of the GFR R&D, several issues are pending, specifically high temperature code qualification of the structural materials especially reactor core and other primary components exposed to temperatures in rage 550 – 850°C. Nevertheless, preliminary screening of the candidate materials is in progress and several Member States are involved. Moreover there is clear synergy with fusion R&D activities, especially plasma facing materials and heat exchanger systems.

The following 3 presentations were made by the IGCAR representatives. The first of which. focused on R&D road of structural materials for FBR in India, specifically Advanced Clad and Wrapper Materials. The current materials to be used for PFBR are D9 steel for oxide type fuel. In future, FM asteels and ODS are considered for cladding and T9 FM steel for wrapper. On-going research regarding the development of new cladding material. primarily focuses on the following issues: (i) ODS alloy composition (ferritic or FM); (ii) processing route (hipping/forging/extrusion), (iii) oxide particle size & distribution and vol. fraction of dispersoid, (iv) microstructural stability – thermal and irradiation, (v) anisotropy and mechanical properties (strength/ductility, toughness), (vi) weldability, (vii) fuel side compatibility, (viii) sodium compatibility and (ix) reprocessing possibilities. An introduction of experimental capabilities as well as an overview of future research outline was briefly summarized. The second .presentation reviewed activity on assessment of advanced core structural materials for radiation damage using ion beams and materials modeling. This initiative is very much linked with ion simulation of radiation damage and application of experimental ion beam facilities. Currently, the IGCAR team is involved .with the commissioning of test facility at 1.7 MeV TANDETRON accelerator in order to develop in house capabilities for simulation of swelling phenomena. As a complementary experimental activity there is also research on application of appropriate compute tools, specifically multi-scale modeling tools to investigate radiation damage at various time and length (space) scale. This presentation reviewed irradiation program at FBTR which is key IGCAR facility in terms of the development and qualification of the fuel and structural materials.

The final .talk of the session summarized technology development for structural integrity evaluation for Japan Sodium-Cooled Fast Reactor. This overview included (i) R&D programme on testing and qualification of structural materials, (ii) design methodologies, (iii) leak-before-break evaluation procedures and (iv) technologies for fitness-for-service including structural reliability assessment methods. It is intended that relevant technologies will be codified in an integrated manner as a set of JSME (Japan Society of Mechanical Engineers) code. The first edition should be published in 2016.

4.5 Coolant Technology and Component Development

Within the frame of the European HELIMNET Network, for both SFR and LFR, a European seminar was organized in France by CEA in order to address the following topics: interaction between coolant and structural material, coolant quality control, handling operations, ISIR, operational and decommissioning issues, severe accidents etc., as well as the main design options. The results of this exercise, aimed to support the strategy of EU with regards the Fast Reactors, have been presented and discussed.

Then ENEA reported the main points related to the comparison of Lead Bismuth Eutectic and Lead behaviour for applications to nuclear systems. Several experimental evidences, gained in recent years, were underlined, and have shown the differences between LBE and Pb as coolants, more particularly in terms of physico-chemistry (O control, purification) and their technological consequences.

The Electra Project, developed by KTH in Sweden, is dedicated to Education and Training for Fast Reactors and more particularly to Heavy Liquid Metal Fast Reactors. The implementation needs to address the main following items: nitride fuel fabrication, reactor design and fuel reprocessing. Then IGCAR has given a detailed overview of thermochemical investigations on Pb-O, Bi-O, LBE-O and Pb-M-O (M: Fe, Cr) systems, and more particularly the data related to oxides, diffusivity, activity and solubility of oxygen in these liquid metals, which are essential parameters to understand the processes that occur in the heavy metal coolant – steel interface. Further, the development of systems to control and monitor the oxygen in liquid metals has been described: a compact design with a YSZ thimble, glass soldered to a metallic component.

CEA presented the R&D strategy related to the ASTRID project: In the framework of the French Act of the 28th of June 2006 about nuclear materials and waste management, a GENIV and actinides incineration demonstration prototype is to be commissioned in the 2020 decade. The paper resumes the R&D program led currently in support to the selection of ASTRID options, particularly on the following topics: core design, decay heat removal system, fuel handling systems, energy conversion systems, etc.. Moreover, the development of a strategy in support to the limitation of core degradation consequences has been described.

Then IGCAR presented the on-going PFBR component test program at Kalpakkam. It includes testing of components such as in vessel and ex vessel fuel handling machines, viz. Transfer Arm and components of Inclined Fuel Transfer Machine (IFTM) - Primary Ramp and Primary Tilting Mechanism, Under Sodium Ultrasonic Scanner, Mutual inductance type Continuous level probes, DSRDM Electromagnet against thermal Shocks, Leak detector layout for PFBR and Gas Entrainment studies in sodium.

IGCAR gave a brief overview of various R&D activities being carried out at IGCAR, and more particularly, for future FBRs, passive / additional safety features which have to be introduced in reactor shutdown systems, so that the overall failure frequency of reactor protection system can be reduced to less than 10-7 per reactor year (compared to 10-6 for PFBR). The development of decay heat removal after shutdown is one of the most important safety functions and must be accomplished with very high reliability. Safety criteria followed for plant design requires that the non-availability of Decay Heat Removal function shall be less than 10-7 per reactor-year.

IGCAR synthesized its important activity in developing chemical instrumentation for in-line measurements in Na, based on electrochemical technologies: hydrogen, carbon and oxygen. In addition, to detect any steam leak in steam generator during start-up and low power operation of the reactor, diffusion based hydrogen detection system has been developed to measure hydrogen in the argon cover gas using a Thermal Conductivity Detector (TCD).

Then IGCAR carried out various experiments conducted on the model steam generator. A 5.7 MWt capacity Steam Generator Test Facility (SGTF) has set up in IGCAR, Kalpakkam to test a 19-tube model steam generator of PFBR. The material of construction, tube dimensions and operating parameters of SGTF SG are similar to that of PFBR SG except the number of tubes.

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Many experiments were successfully conducted on SGTF SG and the results have provided valuable feed back to the design of SG of PFBR and future FBRs.

IGCAR described its developments in progress towards development of Active Magnetic Bearing (AMB) and ferrofluid seal for centrifugal sodium pump. Ferrofluid seals employ a liquid that responds to an external magnetic field and functions as a 'liquid O ring' to achieve sealing. This development will be a very promising step forward for pump technologies.

A description of the developments of Eddy Current Flow Meters for Sodium Flow measurements in FBRs was given, showing all the interest of this technique, illustrated in particular by a measurement carried out in PHENIX.

Finally, the qualification of shutdown systems for PFBR is described: their respective performances are reported and their endurance in high temperature sodium environment was demonstrated. The fourth stage of qualification is the performance evaluation of the system under simulated seismic conditions at room temperature. The commissioning of uniaxial shake tables for giving excitation at multiple support points along with the CSRDM is currently under progress for this study.

4.6 Thermal Hydraulics

This session provides an overview of the experimental and computational works and developments in different countries of the community. Of special interest is the combination of experiment and simulation for the development of new codes as well as for the validation of existing codes on the one hand. On the other hand there are already first examples shown for the feedback of the simulation to the design optimization.

The report on work aims on Uncertainty assessment of core thermal hydraulic analysis for a sodium cooled fast reactor was given by the KAERI specialist. The presentation summarized the achievements on determination of the uncertainty in design parameters, TRU metal fuel performance and transmutation of high level wastes in large scale plant operation. One of the codes used, is MATRA (originally LWR code comparable to COBRA) for sub-channel analysis with mass, energy and momentum transport with adopted heat transfer and pressure drop for SFR. Different direct methods for uncertainty analysis of hot channel factors and the possibility for sensitivity analysis are included, as major sensitivity parameter the inner and outer friction in the fuel assembly are identified. In parallel an uncertainty analysis using Monte Carlo method with randomizing of the input parameters has been established. Comparable uncertainty has been observed in the results for the different codes, even on slightly different absolute values.

A short overview of the results of activities on dynamics and stability analyses of the European LFR demonstrator ALFRED was given by Italian participant. The aim of the research work was to provide preliminary studies for the stability of transients in the core and the primary circuit to support the designer from the onset and to avoid a design without pre-checking of the pre-conceptional design. For this reason, in the early phase, a compromise has to be found between accuracy and efficiency. One zero dimensional model has been derived for studies of the robustness and stability of the system. A second lumped parameter model has been developed for the response to transients with a one dimensional representation of core and primary circuit. Oscillatory behavior at low power level has been discovered, in the case of the consideration of the primary system with a destabilizing contribution from the

steam generators. First transient studies on small perturbations starting from nominal power confirm the robustness of the currently available core design.

A wide overview of the thermal hydraulic studies with different level of detail (1D to 3D) was given by the IGCAR expert. Several important highlights were emphasized, in particular, (i) efficiency assessment of blockage adaptor wholes with 3D CFD analysis and comparison with experiment, (ii) 3D simulation of a full fuel element with 217 rods, (iii) sodium mixing in the plenum above the fuel of the FA. It was also reported that IGCAR contributed to the Monju Benchmark organized by the IAEA. Some interesting examples for the application of CFD for the optimization of the design of coming up systems have been shown. In particular, the validation of the thermal stripping model in the CFD code and the application to different sensitive positions in the whole primary circuit was shown. The results were used for optimizing the structural configuration above the core. CFD studies are used to evaluate methods for the limitation of gas entrainment. For the optimization of the pump, a combination of CFD and comparison with experiments for PFBR and CFBR is applied.

The .final talk of this session brought the overview of actual problems of fast reactor thermal hydraulics which was presented by IPPE representative. The presentation addressed open issues related to the development of engineering solutions for BN-1200 design. . Insight was given into the R&D works supporting the BN-1200 design. Several experiments were conducted or foreseen for validation of codes and optimization of new design features for BN-1200 at IPPE facilities (e.g. elaborate sodium bundle experiments, experiments for vessel thermal hydraulics under steady state, transient and accidental conditions). The experiments are used for the development of a theory for an approximate solution for the vessel thermal hydraulics. Computational modeling of cold trap for Sodium cleaning was done representing the complex combination of thermal hydraulics and dynamics and physical as well as chemical kinetics. The results of the code have been validated on the experiment. New experiments for the sodium-water heat exchanger are under construction and the reactivation of several test facilities for SFR and LFR is foreseen.

4.7 Reactor Safety

The session dedicated to Reactor Safety covered a broad range of subjects, with one common characteristics, i.e. the necessary, permanent interaction between numerical and experimental simulation. The enhancement of safety features for future fast neutrons reactors needs the development of validated computer codes, e.g. for seismic evaluation, code disruptive accident and severe accident analysis and for the design of mitigation systems such as core catcher. The experimental and theoretical analysis is also needed on the side of prevention, like in the case of sodium fires.

The set of small, medium and large size experimental facilities developed in India and results in support of the PFBR and CFBR safety were presented comprehensively, including plans for the future. Worth mentioning are the works dedicated to core catcher technology, decay heat removal systems, seismic evaluation, sodium fire, etc.

On the example of European projects on Generation IV reactors like SARGENIV, it was also recalled all the benefits associated with international collaborative projects, sharing views on safety methodologies and aiming at harmonization of safety standards.

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5. SUMMARY OF PRESENTATIONS

5.1 S. Fomin (KIPT, Kharkov, Ukraine)

The results of further development of prospective conception of fast reactor with the so-called intrinsic safety based on the phenomenon of self-sustained regime of nuclear burning wave [1-4] are presented. We consider a critical fast reactor of cylindrical form with the metal fuel of the mixed Th-U-Pu cycle that consists of two homogeneous zones along the axial direction. In the first zone (ignition zone) the fuel consists of uranium enriched with plutonium. The second zone (breeding zone) is filled with 232Th and 238U. Both zones also include the constructional material Fe and the Pb-Bi eutectic coolant. The initiation and propagation of the nuclear burning wave (NBW) in a fast reactor (FR) are described on the basis of a deterministic approach using the non-stationary multi-group diffusion description of neutron transport taking into account the burn-up of fuel components and the nuclear kinetics for precursor nuclei of delayed neutrons. We present here the calculation results of the space-time evolution of neutron flux and fuel components in such a system for different values of main parameters of the reactor such as its radius, the initial ratio of fuel components (Th-U) in breeding zone and the fuel porosity. The possibility of creating a self-sustained regime of a running NBW in the critical fast reactor under consideration has been demonstrated for the mixed Th-U fuel. The average fuel burn-up of about 50 % can be attained for both Th and U fuel components independently from their initial proportion. The velocity of NBW propagation strongly depends on the transverse reactor size and Th-U ratio. The calculations show a notable stability of the NBW regime towards distortions of the neutron flux in the system and fuel non-uniformity. This stability demonstrates the most important feature of the reactor under consideration, namely, its intrinsic safety ensured by the negative reactivity feedback, which is inherent to the NBW regime.

[1] L.P., Feoktistov. Sov. Phys. Doklady, 34 (1989) 1071. [2] E. Teller. Nuclear Energy for the Third Millennium. Preprint UCRL-JC-129547, LLNL,

1997. [3] H. Sekimoto, K. Ryu, Y.Yoshimura. Nucl. Sci. Engin. 139 (2001) 306. [4] S. Fomin et al. Annals of Nucl. Energy, 32 (2005) 1435; Prog. in Nucl. Energy, 50

(2008) 163.

5.2 D. Sunil Kumar (IGCAR, Kalpakkam, India)

Ferro-boron as a neutron shield is planned for future FBRs to be constructed in India in place of stainless steel boron carbide combination. Ferro-boron has the potential to bring down the cost of neutron shield material inside the reactor as well as from the point of view of volume and weight reduction. This work presents different optimization studies for the objectives of a) keeping the secondary sodium activity below the permissible limit b) increasing the detector counts. Secondary sodium activity depends on the both radial and axial shield configurations. At the same time flux at detector location are also sensitive to the axial shields particularly on the core region. Theoretical analyses with different shield configuration for axial shields in fuel and blanket assemblies done to achieve reduction of shields without reducing detector counts at the monitoring location, for core flux monitoring in PFBR are presented. The reduction of radial shields increases the secondary sodium activity, without affecting the counts, but the reduction of axial shields increases the detector counts and at the same time it will in turn increase the secondary sodium activity. 2D transport calculations done for different axial configurations above Core-1 subassemblies capable of increasing the

detector flux are presented. With this new radial and axial shield configuration the dose at the steam generator goes up slightly as compared to PFBR but is well within the permissible limit. The detector counts actually increase by 54 %. The shield material cost is also lower by a factor 5 as compared to PFBR.

5.3 B. Merk (HZDR, Rozendorf, Germany)

The current work at HZDR will be presented with special emphasis on the development strategy for the DYN3D code as a main component of a diverse 3D coupled core simulation tool for fast reactors and on the work on designable feedback coefficients for sodium cooled fast reactors.

DYN3D is a code for steady-state and transient analysis, currently updated for the use for fast reactors. The code has been extended to multi-group use as well as to the solution of the SP3 equations on rectangular and recently to triangular grid. First verification results for the new triangular multi-group solver will be presented and compared to a HELIOS reference solution. The thermal hydraulics of the code has already been updated with the sodium properties for the steady-state and transient core simulation. In an industry funded project the fuel rod modeling will be improved by coupling with a fuel rod analysis code and by extension of the model to consider fuel rod expansion. First full core tests for SFR will be performed within ESFR. LFR validation will be performed on the Guinevere experiments at Mol/Belgium in the project FREYA. The validation of the code for SFR is foreseen in a cooperation project with the IPPE in Obninsk/Russia, already under negotiation. After these validation projects, DYN3D will be a diverse, well validated 3D nodal code for fast reactor steady state and transient analysis.

The new idea of improving the safety coefficients by the insertion of moderating material will be presented. The effect of moderating material on the sodium void effect, the neutron spectrum, and the kinf is investigated. The use of a zirconium hydride ZrH moderator improves the fuel temperature effect, the coolant effect of the system and the sodium void effect significantly. All changes lead to a significant increase in stability of the fast reactor against transients. The effect of different spatial arrangements of the moderating material is investigated. It is demonstrated that the insertion of the moderating material does not have significant influence on the fuel element power and burnup distribution. The use of fine distributed moderating material creates a new degree of freedom in the design of sodium cooled fast reactors without implying constraints on the core and the fuel element design. It opens the way to create designable feedback effects in a fast reactor core to optimize the response of the reactor core to transients and incidents. The moderating material has only a small influence on the breeding effect and the MA production.

5.4 P. Wolniewicz (Uppsala University, Uppsala, Sweden)

Formation of coolant void can lead to an increase in reactivity in metal-cooled fast reactors. Accordingly, the ability to detect formation of void and similar phenomena is highly relevant in order to counteract transient behaviour of such a reactor. As this work shows, the energy distribution of the neutron flux in a fast reactor is sensitive to formation of void. For monitoring purposes, this fact suggests the use of fission chambers with different isotopic

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content and thus different fission threshold energies. In such a way the monitoring system may be tailored in order to fit the purpose to obtain spectral information of the neutron flux.

In this work, simulations have been performed using the Monte-Carlo-based code SERPENT on the ELECTRA reactor design, a 0.5MWth Lead-cooled Fast Reactor (LFR) planned for in Sweden. The simulations show significant changes in the neutron spectrum due to the formation of void located in specific in-core regions as well as due to a homogeneous core-wide distribution of small bubbles. In an attempt to quantify and to put a number on the spectroscopic changes, the number of neutrons in the high energy region (2-5 MeV) are compared to the number of neutrons in the low-energy region (50-500 keV) and the changes caused by the introduction of void is analyzed. The implications of the findings are discussed.

5.5 F. Klassen (NRG, Petten, The Netherlands)

Fast reactor systems, cooled by sodium or by heavy liquid metal, are envisaged to play an important role in the future of nuclear energy production because of their potential to i) efficiently use uranium and ii) reduce the volume and lifetime of nuclear waste. The development of fast reactor systems still requires a significant effort, not the least on the part of the fuel. Whereas currently MOX fuels are mostly foreseen for fast reactors, for the future also other fuels are envisaged. Generally, new fuels offer improved economics by making more efficient use of resources (e.g. thorium), provide additional safety margins, or provide an option to reduce the lifetime of high level nuclear waste, by dedicated recycling of long lived actinides.

When developing innovative nuclear fuels, the behaviour during irradiation needs to be known in detail. Fuels need to fit many requirements, regarding neutronic properties, chemical stability, thermal conductivity, margin to melt and reprocessibility. These aspects can only be fully assessed in dedicated test irradiations. At the High Flux Reactor (HFR), a 45 MWth materials test reactor in Petten in the Netherlands, numerous irradiation experiments have been performed to assess the behaviour of innovative fuels for fast reactor applications. These include fast reactor MOX fuels, nitride fuels, inert matrix fuels (IMF), thorium-containing fuels, as well as Minor Actinide (MA) containing fuels for transmutation purposes.

.

In this paper, an overview will be given of the recent activities at the HFR regarding advanced fast reactor nuclear fuels, with respect to fabrication, irradiation and post irradiation examinations (PIE). Focus will be on innovative, plutonium containing fuels and on transmutation fuels.

To provide this overview, a number of recent irradiation projects will be described.

- BODEX focuses on the investigation of helium release mechanisms and impact by a controlled helium generation during irradiation in boron-doped inert matrices;

- CONFIRM involves the irradiation of (Pu,Zr)N at high linear power;

- MARIOS & SPHERE for both homogeneous and heterogeneous recycling, with a comparison of standard pellet-type fuel with sphere-PAC.

In ‘heterogeneous’ blanket fuel, for instance the (U,Am)O2-x proposed for sodium-cooled fast reactors, concentrations of minor actinides and corresponding helium production are much higher compared to ‘homogeneous’ fuel (Pu,U,Am)O2-x, but the irradiation conditions are milder. These irradiation programmes are currently underway, and aim at providing insight towards the optimal recycling of minor actinides in a sodium-cooled fast reactor.

5.6 T. Satyasheela (IGCAR, Kalpakkam, India)

Static and dynamic studies of metal fuelled fast breeder reactors (MFBR) are carried out to verify the passive shutdown capability and its inherent safety parameters. Static calculations are carried out to determine the vested reactivity feedback parameters from the fuel temperature and coolant temperature raise separately. Inherent safety and the maximum coolant temperature of the reactor are calculated based on the reactivity feedback parameters. It is found, reactivity decrement of metal fuel reactor is small as compared to oxide fuel. It is concluded that, metal fuel can go to sub-critical state with comparatively smaller negative feedback reactivity during ULOFA, and its maximum coolant temperature is below its saturation point. In case of UTOPA, outlet coolant temperature is less than its saturation point with both the metal and oxide fuel

Unprotected loss of flow [ULOF] analysis of metal (U-Pu-6% Zr) fuelled 500 MWe and 1000 MWe pool type MFBR are studied with a flow halving time of 8 S. Study is also made with uncertainties (typically 20 %) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident [ULOFA] and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80 % core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20 % uncertainty on the sensitive parameters.

Analyses are carried out with a higher flow halving time such as 15 s and 30 s. With the extended flow halving time, even with 20 % uncertainties on sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity. From the study it is concluded that, for ensuring the safety of 1000 MWe MFBR core, with considering the uncertainties, either a reduction of Na void reactivity effect or increase in flow halving time is required. Higher primary pump flow halving time (15 s instead of 8 s) can avoid cliff edge effects in 1000 MWe MFBR transient behaviour.

5.7 D.K. Mohapatra (IGCAR, Kalpakkam, India)

It is a well known fact . that metal-fuelled fast breeder reactors (MFBR) will be introduced in India’s nuclear power program to enhance the fuel-breeding ratio, consequently lowering the

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reactor doubling time. In this connection, several reactor designs with varied core composition have been suggested. The metal fuels considered in the MFBR design are alloys of Uranium-Plutonium-Zirconium with varied concentration of zirconium. In this study, the physics design aspects of a 1000 MWe MFBR are elucidated and the core neutronics analysis results for the same have been presented. Apart from the neutronic parameters, the safety parameters of the MFBR have also been predicted for further safety analyses. To understand the fuel burnup effects in a metal fuelled fast reactor system, systematic burnup analysis is carried out for a 500 MWe MFBR design and the results are compared with that of a similar MOX fuelled FBR. The results indicate that in a metal core the loss of reactivity with burnup is only half that of the oxide core. Interestingly, the breeding ratio remains nearly constant in metal core, but increases with burnup in MOX core. It is also found that isotopic composition of Plutonium is more stable with burnup in metal core compared to that in MOX core. Since multiple fuel recycling is interlinked with FBR growth, this behavior indicates that it is perhaps easier to maintain constant Plutonium isotopic composition (Pu vector) with multiple recycling in metal-fuelled FBR as compared MOX fuelled ones. Further, thorium has been introduced in the blankets of the MFBR under consideration and the effects on the core parameters are studied. Thorium introduction is analyzed for different blanket configurations. It is observed that by introducing thorium in the MFBR, the total breeding ratio decreases. However, the internal breeding ratio remains more than unity, which means plutonium production in the core is self sustained.

5.8 G.Raghu Kumar (IGCAR, Kalpakkam, India)

Fast Breeder Test Reactor (FBTR) is, India’s first Sodium Cooled fast reactor, fuelled with mixed carbide. As there was no sufficient experience with the carbide fuel, a small core with 70% of PuC and 30% UC (called as MK I fuel) with 27 fuel sub assemblies was planned. This core was planned to operate at peak linear power of 250W/cm and peak target burn up of 25,000 MWd/t. Subsequently it was decided to go for full core of 85 SA with 55% of PuC and 45% UC (called as MK II Fuel). During the process of conversion from small carbide core to nominal MK II core 13 MK II fuel SA are progressively added. At one stage we planned to convert the core to hybrid core driven by carbide fuel surrounded by MOX fuel with 44% of PuO2 and eight MOX sub assemblies were added. Presently the core consists of 27 MK I, 13 MK II and 8 MOX type SA, with various burn up (BU). Also different fuel sub assemblies had varying Plutonium vector.

During the refueling at the end of the every irradiation campaign, the change in the reactivity due to fuel handling was being calculated using two dimensional diffusion R-Z code ALCI, with 25 group Cadrache cross section set. As core became complicated it was becoming difficult to model the core in the two dimensional R-Z code also in the fourteenth campaign lot of fuel changes were planned and it was thought that it would not be possible to predict the change in k-eff to the desired accuracy of 150 pcm as prescribed in the Technical Specifications of FBTR. Due to this, the fuel handling operation was done in four stages. Subsequently it was decided to use Monte Carlo code MONICA. This code is tested against MCNP and the input specifications are same for both the codes.

In order to avoid manual errors in the preparation of input, an EXCEL Work Sheet comprising of different MACROS was made, which creates “Input” for the Monte Carlo code. The input takes care of axial burn up in each of the fuel sub assemblies. The input for the code is tested for all the campaigns. In this paper we present the functions of the work sheet macros, in exact core modeling, with appropriate Plutonium vector and varying axial burn up. The paper also gives the results of the calculations of K-eff in various campaigns and

its comparison with the measured value. The paper also gives the proposed improvements to the MACROS.

5.9 P.Le Coz (CEA, France)

In the framework of the French Act of the 28th of June 2006 concerning. nuclear materials and waste management, a GEN-IV and actinides incineration demonstration prototype is to be commissioned in the 2020 decade.

This prototype, called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) sets out to demonstrate the progress made in SFR technology on an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. It will also be used as a test bench for advanced inspection and repair techniques. More specifically, we are aiming for a level of safety that is at least equivalent to that of the third generation reactors and for the integration of the feedback of the Fukushima accident, with progress made in SFR-specific fields. ASTRID will also be designed to investigate and to demonstrate the feasibility of transmutation of radwaste: It will be replacing the Phenix plant in terms of an irradiation reactor so as to test homogenous and heterogeneous minor actinide recycling modes, whose industrial interest are currently being assessed in the frame of the above mentioned Act.

The first deadline set by the French Act of 28 June 2006 is for late 2012 at which time the authorities must be provided with an assessment of the industrial perspectives on actinides transmutation. To remain consistent with the general schedule fixed by the Act, the CEA decided to launch the phase 1 of the preliminary design of the ASTRID prototype in 2010 in order to . present at the same occasion (late 2012) preliminary budgetary evaluations (especially in terms of amount and investment schedule), and technical facts so that a decision on the following steps can be reached. An assessment of the fixed and still open-ended options will be established at this date, together with the planning and cost evaluation (including R&D) of next phases. Phase 1 of the preliminary design also aims at obtaining an initial opinion from the French Nuclear Safety Authority (ASN) on the safety orientations recommended for ASTRID, in order to check their suitability.

Phase 2 of the preliminary design will take place between 2013 and 2014, including the drafting and submittal to ASN of the Safety Options Report, a deeper assessment of retained options and hard points, and a consolidation of cost and planning figures. In order to benefit from the unique feedback existing in France on Sodium Fast Reactors, the CEA partnered, among others, with EDF, as experienced SFR operator and general architect, and with AREVA, as experienced SFR Nuclear Island engineer and component’s designer. CEA is still willing to conclude other partnerships with French or international companies. This paper presents the current state of the project, the first design options and the main milestones

5.10 D. de Bruyn (SCK•CEN, Mol, Belgium)

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK•CEN in replacement of its material testing reactor (MTR) BR2. The Belgian Federal Government has

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approved early 2010 the funding of this international project, which from 2023 onwards, will contribute to the development of innovative solutions in the field of nuclear technologies.

The MYRRHA-facility, currently developed with the aid of the FP7-project ".Central Design Team" is conceived as a flexible irradiation facility (100 MWth), able to work in both subcritical (ADS) and critical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications and finally allow also Si-doping.

MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at a reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on the heavy liquid metal technology, the eutectic lead-bismuth, it will be able to significantly contribute to the development of Lead Fast Reactor Technology and in critical mode MYRRHA will play the role of European Technology Pilot Plant in the roadmap for LFR.

In this first presentation special attention was paid to the latest configuration of the reactor core and primary system. In another presentation foreseen during the meeting, the present plant layout will be presented together with the historic evolution and the international positioning of the MYRRHA-project.

5.12 P. Puthiyavinayagam (IGCAR, India)

.Commercial Fast Breeder Reactors (CFBR) would incorporate several advanced design features towards economy. In the domain of reactor core, the design objectives are high performance characterized by higher burnup, higher cycle length, improved breeding, increased core average coolant temperature, reduced nuclear material specific inventory and reduced throughput to reprocessing plants given the closed fuel cycle program that India has adopted. Further, design optimization is also pursued towards simple manufacturing features, optimum inspection requirements coupled with improved and rationalized design improvements.

The drivers for core design improvements are economy, safety and reliable operation. The basis for such an exercise stem from the operating experience from FBTR, design, safety review and manufacturing experience from PFBR while keeping in mind the strategic considerations on the infrastructure. This led to identification of some design parameters for optimization towards economy. The major parameters are fuel pin neutronic and thermal hydraulic design and choice of clad wrapper materials which are coupled with respect to their performance. The following are the key requirements for high performance and economy which are currently focused on for further study: (i) development and rigorous characterisation of fuel and clad materials (ii) generation of PIE data across entire data ranges from FBTR and also from the initial performance of PFBR core (iii) understanding of fuel behaviour by accurate modeling (iv) optimum thermal hydraulic design and (v) design optimisation, standardisation and consolidation.

This paper brings out in brief the approach adopted and few design improvements that are currently undertaken for CFBR core design. The material, safety aspects and metal fuel development aspects are not covered in this paper.

5.13 V. Rajan Babu (IGCAR, India)

Prototype Fast Breeder Reactor (PFBR) which is a 500 MWe, mixed oxide fuelled & sodium cooled reactor, is under construction at Kalpakkam. Further, Department of Atomic Energy has planned to construct six more FBRs of 500 MWe size by 2020. After the techno-economic demonstration of FBR technology through PFBR, it is essential to achieve high economic competitiveness along with enhanced safety on par with other power generation options. In this regard, cost effective measures have been identified and worked out in detail for future FBRs covering all aspects of the reactor with particular focus on reactor assembly components design. In this context, innovative concepts such as welded grid plate with reduced height, inner vessel having single toroidal shaped redan integrated with fuel transfer post, dome shaped roof slab with optimum support location to minimize the seismic moments, thick plate rotatable plugs, safety vessel along with thermal insulation integrated with vault liner and increased number of primary pipes have been conceived for future FBRs. The design features of these components and the technology development activities undertaken for the critical long delivery components are detailed in this paper.

5.14 R.Vijayashree (IGCAR, Kalpakkam, India)

The environmental pollution caused by the release of green house gases by fossil power plants, associated global warming and also depletion of all known sources of fossil fuels are the factors that are driving the world to look for alternate sources of power generation. For the nuclear reactors to gain their legitimate role as a clean source of power, the possibility of major accidents (like that of Three Mile Island in 1979, Chernobyl in 1986 and Fukushima in 2011) is one of the major obstacles. Hence, the augmentation of nuclear reactor safety is very essential for any reactor to gain public acceptance and global importance. The shutdown system and decay heat removal system are the two important subsystems that ensure the nuclear reactor safety. The reliability of the shutdown system can be considerably enhanced by using diverse and passive principles for shutting down the reactor.

Any shutdown system comprises of a reactor protection system and Actuation system. The sensors signal processing units and logic are under the reactor protection system. The absorber rods and their drive mechanism form the actuation system. .The reliability analysis of the existing configuration of the shutdown system of PFBR has indicated that they are contributing almost equally to the failure of shutdown system.

Passive means of triggering the introduction of neutron poison in the system on global faults helps to enhance the reliability by augmenting to that of reactor protection system. However this alone is not sufficient to enhance reliability because the unreliability due to actuation system starts dominating. Hence it is proposed to introduce a third shutdown system based on liquid poison/absorber balls having both active and passive triggers for the introduction of poison into the core on global faults In addition the existing design of electromagnet of Diverse safety rod mechanism is enhanced to that of temperature sensitive Electromagnet. Stroke limiting devise which is an engineered safety feature is also proposed to be introduced in the existing design of Control and Safety Rod Drive Mechanism. This paper presents a brief description of the above mentioned enhancement proposed in the existing design of shut down system. The conceptual design of third shutdown system under consideration is also presented.

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5.15 S. Raghupathty (IGCAR, Kalpakkam, India)

Systematic efforts have been made worldwide to reduce the capital cost of fast breeder reactors in order to make them economically competitive with light water reactors. Fast breeder reactors are expected to make a major contribution in future in both the second and third stages of the Indian Nuclear programme to meet the energy requirements of India. After the successful construction and operation of Fast Breeder Test Reactor (FBTR), construction of Prototype Fast Breeder Reactor (PFBR) is under progress. Beyond PFBR, it is planned to construct six more 500 MWe units and 1000 MWe metal fuelled reactors thereafter.

The next two units of 500 MWe capacity called ‘Commercial Fast Breeder Reactor (CFBR)’ is planed as twin units at Kalpakkam adjacent to PFBR. Fuel handling systems account for ~ 4-5% of the total capital cost. Being a twin unit and fuel handling being carried out off-line, there is . merit in sharing most of the fuel handling system between the two units to achieve improved economy. Also, design simplifications are planned to further improve economy and safety. This paper brings out the highlights of the studies carried out including details of a innovative twin unit layout with sharing of fuel handling systems and outlines the design approach for future FBRs.

5.16 V. Balasubramaniyan (IGCAR, Kalpakkam, India)

Successful operation of a 40 MW(t)/13 MW(e) capacity fast breeder test reactor over 25 years, strong R&D executed in multidisciplinary domain and construction of a 500 MW(e) prototype fast breeder reactor (PFBR) based on an indigenous design have provided high confidence in the success of fast breeder technology in India. Beyond the PFBR, there are plans to construct six more FBRs, each of 500 MW(e) capacity. Towards this end, a systematic roadmap has been drawn up for improved economy and enhanced safety through a number of measures. These include major features incorporated to achieve economy such as the twin unit concept, plant life increased to 60 years in comparison to 40 years for the PFBR, reduced fuel cycle cost with higher burnup, minimizing the use of SS 316LN for NSSS components, reduction in special steel specific weight requirements, improved load factor, etc.

Consistent with the above approach and also keeping in view the significance of heat transport systems to plant economics, overall safety and reliable operation, several innovative design features and methodologies are examined for future FBRs. These include choice of ferritic steels for sodium piping & valves, optimization of design and number of steam generator modules with reduced number of tube to tube sheet joints, in-vessel primary sodium purification, optimization of number & enhanced diversity in the decay heat removal system etc. The paper brings out the highlights of the studies and summarises the choices for future FBRs.

5.17 K. Madhusoodhanan (IGCAR, Kalpakkam, India)

Based on the experience in designing PFBR, some design changes are proposed, which will help in improving economy. Some of these are either new techniques or design improvements. Major activities are mentioned below.

Development of core thermocouple probe with three thermocouples: in PFBR, Fuel Subassembly outlet temperature is monitored by a probe consisting of two thermocouples. Signal processing is by three computers, which causes a agreat deal of cross wiring.

Probe design is modified to accommodate three thermocouples. A computer processes signal of each thermocouple. This will simplify the core temperature monitoring system and make it in line with the safety criteria (independent and triplicate instrumentation for safety class-1). Development of signal processing electronics suitable for placing in the roof slab top will reduce the requirement of large volume of signal routing from roof slab to the periphery. The difficulty in finding independent and triplicate cable routes will be solved by this development. Large quantity of connectors can be eliminated.

Wireless instrumentation for Roof slab area: wireless based technology is adopted for some of the areas where space constraints are faced. Laying of cables and installation of disconnectable cables (whose volume is very large) is difficult. This technology will help in reducing the size of trailing cable system, or it can be eliminated in total.

Sodium level probes: mutual Inductance level probes used in PFBR are very long for most of the tanks. It poses difficulty in manufacturing as well as handling at site. Also the difficulty to provide support at bottom against flow induced vibration forced to reduce the length there by reducing the measurement range. Level measurement based on RADAR principle, which is available for other fluids are tested for sodium and found promising. In this case, a small antenna is installed at the top of the tank. Efforts are being made to gain confidence in the usage of RADAR level probes for sodium. This will avoid the problems associated with mutual inductance level probe.

Development of neutron detectors: high temperature Fission Chambers (HTFC) with 0.2 cps/nv sensitivity for application in high temperature (567 C) of fast reactors are developed indigenously and are under testing. These detectors are for application in control plug. For deploying under the safety vessel, fission chambers with 0.75 cps/nv sensitivity is being developed. These detectors together are covering the full range of operation for PFBR. For future reactors, the 0.75 cps/nv under vessel detectors will be qualified for high temperature (567 C). This will be located in control plug and will take care of full range of operation. Under vessel detectors are eliminated from design.

Qualification of Indian manufacturers of thermocouple: efforts are being made to qualify thermocouples made by Indian industries as per the requirements of PFBR. This will standardize the products and reduce the cost of testing.

Development of leak tight penetration assemblies for electrical and I&C cables: development activities are initiated to manufacture through Indian industries.

5.18 S.B. Degweker (BARC, Mumbai, India)

Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. In India, there is an additional source of

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interest in ADSs which is related to utilization of our large deposits of Th. We present studies on various options for Th utilization by the ADS route. Breeding of U233 in fast and thermal spectrum ADSs, for use in critical reactors, is discussed as also the option of power production in ADSs fuelled with Th. We consider heavy water moderated reactors and fast spectrum reactors operated as sub-critical reactors driven by an external spallation source for the purpose of these studies.

In the breeding option we show that it is possible to produce power for driving the accelerator and even some surplus to be exported to the grid without significant loss in the U233 production. In the power production option using ADSs, we consider Th utilization in a once through Th cycle as also the possibility of recycling. In heavy water moderated reactors, the once through cycle represents an attractive ADS concept for Th utilization but the gain is rather low, requiring a high power accelerator and necessitating a substantial fraction of the generated power to be fed back to the accelerator. The self sustaining Th-U cycle in heavy moderated ADSs is a more realistic option, since it has a good gain. In fast spectrum sub-critical reactors, the once through Th cycle has an acceptable gain but requires an initial fissile seed (Pu) to start the system and also requires development of a fuel that can withstand very high neutron fluence.

A sub-critical facility driven by D-D and D-T neutron sources for carrying out ADS related experiments is expected to become operational next year. The experimental programme includes among others, proposals for measurement of the degree of sub-criticality by various deterministic and noise methods. We discuss activities and developments related to the methods for measuring the sub-criticality of ADSs. This includes a new theory of noise in ADSs with a non-Poisson source, development of a noise simulator and studies on location of detectors for this purpose.

5.20 T. Jayakumar (IGCAR, Kalpakkam, India)

For India, nuclear energy is an inevitable option for supply of large amount of clean energy. Towards this, a three-stage programme is conceived to exploit fully the indigenous uranium and thorium resources through water reactors and sodium-cooled fast reactors (SFR). Successful operation of 40 MWt/13 MWe capacity fast breeder test reactor (FBTR) over 25 years, construction of 500 MWe prototype fast breeder reactor, which is in the advanced stage of construction, strong science based R&D executed in multidisciplinary domains have provided high confidence for the success of SFRs. Establishing techno-economic viability through SFRs with mixed uranium-plutonium oxide fuel and closed fuel cycle, design for long life nuclear reactors (100 years), significant reduction of capital cost and fuel cycle cost by targeting 200 GWd/t burnup of fuel, metallic fuel with high breeding ratio to achieve shorter doubling time (< 10 y), construction of a series of large size reactors (1000 MWe) with co-located fuel cycle facilities by adopting mega park concept are important priorities for India. Realisation of these priorities would demand a variety of high performance materials for in-core components, reactor structurals and steam generators. IGCAR has made a challenging roadmap for indigenous development of these materials, along with material cycle management aspects such as thermo-mechanical processing, fabricability, mechanical property evaluation, damage mechanics, structure-property correlation and structural integrity.

Materials and operating environments for core structures in SFRs include intense neutron radiation necessitating materials resistant to void swelling and irradiation embrittlement/creep, and elevated operating temperatures for structural materials, necessitating incorporation of creep and low-cycle fatigue properties in design. Present SFRs use 316LN austenitic SS as major construction material of reactor components, irradiation-induced void swelling resistant 15Cr15Ni2.2Mo-Ti austenitic SS Alloy D9 as material for fuel clad and wrapper and modified 9Cr1Mo steel in steam generators. For higher fuel burnup in future SFRs, improved fuel clad materials being developed include IFAC-1 (Indian Fast Reactor Advanced Clad-1, similar to Alloy D9I with higher phosphorus and silicon) and 9Cr oxide-dispersion strengthened (ODS) ferritic-martensitic steel. Also, plain 9Cr-1Mo steel for the wrapper and modified 9Cr-1Mo steel for the cladding are being developed for SFRs with metallic fuels. Other improved materials being developed for the future SFRs include nitrogen-enhanced 316LN SS for enhancing the creep strength of reactor components, and type-IV cracking resistant boron-added modified 9Cr-1Mo steel for the steam generators. Materials being studied also include ferro-boron, cobalt-free hardfacing alloys and magnetic switch materials, as also welding consumables for all the advanced materials. This paper highlights the work and approaches for the successful deployment of materials for the current and future SFRs.

5.21 B.K. Panigrahi (IGCAR, India)

Radiation damage to the core structural materials, due to intense neutron bombardment, is one of the crucial issues in the context of nuclear energy programme. Radiation environments in the hybrid reactors like ADSS and advanced fast and fusion reactors, that are to be developed in the near future, are currently not available for testing materials. The irradiation environment in these reactors will be complex and requires in-depth understanding of the role of defects and their evolution in order to predict the material behaviour in in-service conditions. Despite a worldwide effort, predicting behaviour of materials in irradiation environment is still a huge challenge. This is primarily due to the multi-scale nature of radiation damage process, and the inherent non-linearity of the underlying processes. The key to understanding these phenomena is to integrate multi-scale multi-physics modeling, and controlled experiments based on sharply focused questions emanating from the underlying issues. Multi-beam accelerated charged particle irradiation experiments are ideal for conducting modeling oriented & parametric studies with rapid feedback as the damage rate with ions from accelerators are typically 103 times higher as compared to irradiation experiments using neutrons from fast reactors. With simultaneous multiple beam irradiation, it is possible to simulate synergistic effects of displacement and transmutation introduced by high-energy neutrons. Also, it is easy to setup in-situ experiments for investigating point defects dynamics and to explore long term kinetic pathways.

A strong R&D programme for the study of radiation damage using charged particles from ion accelerators is being pursued at IGCAR in order to obtain comprehensive understanding of radiation damage phenomena with particular reference to fast reactor materials. A unique experimental facility for measurement of void swelling at damage levels of more than 150 dpa using heavy ion beam from 1.7 MV accelerator was set up at IGCAR. Using these facilities, detailed studies on the temperature and dose dependence of void swelling and the effect of minor alloying elements like Ti ,P on the swelling behavior of D9 alloy, the material for clad and wrapper in the FBRs, have been carried out. These studies have played a crucial role in

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arriving at the optimal composition of these elements in a modified D9 alloy (D9I), for achieving a target burn-up of 150 dpa..

A program on multi-scale modeling of radiation damage using high performance computing cluster built in-house has also been initiated. Some of the recent efforts in this area include study of radiation damage cascade using Molecular Dynamics, defect evolution using kinetic Monte Carlo simulations in model materials, and ab initio DFT calculations aimed at understanding the role of minor alloying elements and vacancies in bcc Fe which have important implications in the context of design of nano-structured ferritic alloys, the next generation core structural materials. A significant result of the computation is that the strong oxygen-vacancy interaction leads to increase in vacancy concentration by 16 orders of magnitude in the form of O-V complexes. These ferritic alloys contain a fine dispersion of Y-Ti-O nanoclusters for improved mechanical and radiation resistant properties. The ab-initio calculations show that the binding energy of Y-Ti-O nanocluster increases when Ti is replaced with Zr, which could lead to finer dispersion of nanoclusters resulting in improved performance of ferritic alloys. Kinetic Monte Carlo simulations of kinetics of spatial and size distribution of yttria precipitates in bcc Fe, using a code developed in-house, showed that presence of Ti leads to refinement of the size of the precipitates. The experimental studies on three different model ODS alloy (Fe - 0.3Y2O3, Fe - 0.3Y2O3 - 0.2Ti) and (Fe - 14Cr - 0.3Y2O3 - 0.2Ti) prepared in the form of a rod by ball milling and hot extrusion showed that in presence of Ti, a drastic reduction was found in particle size of ODS alloy and Cr addition did not show a large effect on the size distribution of Yttria dispersion. Further, irradiation studies on model ODS alloy Fe-0.3 % yttria showed an increase in the average Yttria particle size increases from 31. nm of the pristine material to 3.5 and 9.7nm for irradiation condition of 25dpa and 80 dpa respectively indicating that these particles are stable upto 25 dpa. However, at 80 dpa, yttria particles were found to be unstable as evidenced from the increase in average particle size and decrease in particle density. This is in contrast to the reported ion irradiation results in 9Cr ODS steel at 600°C, which showed an increase in particle size and decrease in oxide density. Further work is in progress to understand the role of Ti and other elements in the radiation stability of the oxide particles. The talk will give an overview of these studies.

5.22 V. Kathik (IGCAR, India)

Austenitic stainless steel (AISI 316) is used as the structural material of the Fast Breeder Test Reactor (FBTR) at Kalpakkam. While the core components such as cladding and wrapper of FBTR are subjected to high fluence neutron irradiation, the permanent core structurals like reactor vessel, thermal shields, grid plate and other block pile components experience low fluence irradiation conditions over their lifetime. Fast neutron irradiation in the structural materials results in changes in mechanical properties which are closely linked to the microstructural changes. The performance of SS 316 has been systematically studied through mechanical property evaluation and microstructural analysis at various displacement damages right from 1 dpa up to about 80 dpa. For mechanical property evaluation, the mechanical testing are performed on samples extracted from clad / wrapper tubes of the irradiated fuel assemblies and also on prefabricated specimens irradiated in the FBTR core as a part of planned irradiation experiments.

Tensile tests are carried out on specimens of irradiated cladding, wrapper and prefabricated sub-size tensile specimens as per the ASTM standards remotely in hot cell facility. Special

techniques have been adopted for remotely gripping the thin specimens during testing. In addition, shear punch test technique employing small specimens (8.0 mm diameter and 1.0 mm thick) are also used for evaluating the mechanical properties and microstructural analysis of irradiated wrapper.

The 20% cold worked cladding, irradiated at temperatures of 430-480 C, revealed a decrease of the Ultimate Tensile Strength (UTS) at displacement damages > 60 dpa both in high temperature (480 C) and room temperature (28 C) tests, while the wrapper irradiated at temperatures of 400 C-430 C exhibited a hardening behaviour. The differences in the mechanical properties of irradiated cladding and wrapper could be linked to their swelling behaviour. The softening behaviour of cladding or hardening behaviour of the wrapper was associated with a loss of ductility in both the cases. Shear punch tests of wrapper also indicated a progressive increase in the room temperature yield strength (YS) and UTS with increasing dpa and a decrease in the ductility, consistent with tensile test results. The evaluation of mechanical properties of core materials like cladding/wrapper was very useful for progressively enhancing the peak burnup of the FBTR fuel to 155GWd/t corresponding to a peak displacement damage of 80 dpa.

Towards assessing the irradiation induced changes in mechanical properties of permanent core structure of FBTR after low dose irradiation, an accelerated irradiation test of grid plate specimens (annealed SS 316) were carried out in FBTR. The tensile test results of grid plate material irradiated to 1-2.5 dpa at 350 °C showed radiation hardening accompanied with ductility loss. A uniform elongation of above 20% at test temperatures of 28°C, 350°C & 400°C indicated retention of adequate ductility in SS 316 grid plate of FBTR for an accumulated fast neutron dose of 2.5 dpa.

The salient results of post irradiation mechanical testing out on irradiated SS 316 cladding, wrapper and grid plate materials of FBTR at different fluence levels will be presented in this paper.

5.23 T. Asayama (JAEA, Ibaraki, Japan)

This paper summarizes the ongoing efforts on technology developments for the Japan Sodium Cooled Fast Breeder Reactors (JSFR) focusing on the development of structural materials and design methodologies. JSFR’s operating temperature is 550C and design life is 60 years. Two kinds of new materials, 316FR stainless steel and Modified 9Cr-1Mo steel, are to be applied to main components. The former is low-carbon nitrogen-added 316 steel which has superior creep properties and will be used for the reactor vessel and internal structures. The latter will be applied to most of the coolant systems including primary piping, intermediate heat exchange, secondary piping and steam generator, taking advantage of an excellent combination of thermal conductivity and elevated temperature strength. Research and development including long-term material tests and structural model testing is being carried out for the development of material strength standard and design methodologies that are applicable to 60-year design. Moreover, in order to fully take advantage of the conditions in which fast breeder reactor components are operated, technologies for the establishment of a leak-before-break evaluation procedure and a fitness-for-service code are also being extensively developed. Finally described is a path forward to materialize a systematically inter-related codes and standards structure that allows margin optimization for future fast breeder reactors.

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5.24 C. Latge (CEA, France)

Sodium-cooled fast reactor (SFR) and Lead-cooled fast reactors (LFR) nuclear energy systems are among the six candidate technologies selected in the Generation IV Technology Roadmap for their potential to meet the new technology goals to improve safety, economic competitiveness, inspection and repair and proliferation resistance. SFR is the reference concept for Europe and LFR, with GFR, an alternative option. Within the frame of the European HELIMNET Network, a European seminar is organized in France by CEA, beginning of October 2011, in order to address the following topics, such as interaction between coolant and structural material, coolant quality control, handling operations, ISIR, operational and decommissioning issues, severe accidents etc as well as the main design options.

The results of this exercise, aimed to support the strategy of EU with regard to the Fast Reactors will be presented and discussed.

5.25 A. Gessi (ENEA, Italy)

The goal of this work is to compare critically LBE (Lead-Bismuth Eutectic) and Pb, as coolants for GenIV fast reactors. The choice of Heavy Liquid Metals for a nuclear fast reactors, comes from several known advantages, both technological and nuclear.

Hystorically, LBE was the first choice, due to its very low melting point (125°) compared with Pb (327°C).

However, several experimental evidences, gained in recent years, suggest the need for a deep analysis and comparison between LBE and Pb as coolants, especially as far as technological issues are concerned.

This work is a comparison of the two, starting from basic properties and going through non metallic elements behaviours, (i.e. Oxygen), corrosion, of structural materials and related technologies.

Keywords: Pb, LBE, Oxygen, corrosion, coolant chemistry

5.27 R. Ganesan (IGCAR, India)

Liquid lead and lead-bismuth eutectic (LBE) alloy are currently being explored for service as spallation target and coolant in accelerator driven systems (ADS) and as a candidate coolant in advanced nuclear reactors [i]. Though pure lead and LBE are corrosive towards structural steels, this corrosion can be mitigated by controlling the oxygen concentration in the liquid metal such that a protective oxide coating is formed over the structural material [ii]. An understanding of the thermochemical behaviour of oxygen in liquid lead, bismuth and LBE is needed for the oxygen control in the coolant. Further, the oxide that coexists with LBE saturated with oxygen is not unequivocally known. Thermochemical data of oxides, diffusivity, activity and solubility of oxygen in these liquid metals are essential parameters to

understand the processes that occur in the heavy metal coolant – steel interface. Further, control and monitoring of oxygen in liquid metals is also important. Towards this direction, detailed thermochemical investigations have been carried out in Chemistry Group, IGCAR on Pb-O, Bi-O, LBE-O and Pb-M-O (M: Fe, Cr) systems. Towards the fabrication of the oxygen sensor for Pb and LBE loop, a compact design wherein a YSZ thimble, glass soldered to a metallic component had been chosen.

Activities, solubilities and diffusivities of oxygen in lead, bismuth and LBE were determined by solid electrolyte based EMF techniques [iii]. Partial phase diagrams of Pb-Cr-O and Pb-Fe-O systems have been established by phase equilibration studies. Thermochemical data such as Gibbs energy of formation, enthalpy of formation and heat capacity of various ternary oxides were measured [4, iv,]. Based on the phase diagrams of Pb-Cr-O and Pb-Fe-O systems and the thermochemical data of the ternary oxygen compounds, the nature of passive oxide layer formed over the structural steel has been assessed. The measured oxygen potentials in the phase fields of Pb-Fe-O and Pb-Cr-O show that at temperatures below 900 K, Fe3O4 only would be formed on surfaces of structural steels at oxygen potentials that are maintained in liquid Pb systems. ‘PbFe5O8.5’ could be formed at high oxygen concentrations. Presence of chromium in steels can lead to the formation of (Fe,Cr)3O4 also.With a view to develop oxygen meter for lead and LBE systems, nanocrystalline yttria stabilized zirconia (YSZ) powders have been synthesized by a novel combustion method and these powders could be sintered to high densities and the products exhibited a leak tightness of 10-9 S cc/s to helium, which conform to any device application. Suitable glass, which is compatible with Pb and LBE and having a sufficient coefficient of thermal expansion between ferritic steel and partially stabilised zirconia (PSZ) was chosen based on the data in literature. The sealing conditions were optimized and the sealing was found to be intact after exposure to liquid LBE at 723 K for 500 h

[v].Details of these research activities will be discussed in the presentation.

[1] T. Obera, T. Miura, H. Sekimoto, Prog. Nucl. Energy 47 (1-4) (2005) 577-585.

[1] J. Zhang, N. Li, J. Nucl. Mater. 373 (2008) 351-377.

[1] R. Ganesan, PhD Thesis, Indian Institute of Technology Bombay, Mumbai (2006).

[1] S. K. Sahu, PhD Thesis, University of Madras, Chennai (2011).

[1] S. K. Sahu, R. Ganesan, V. Jayaraman, T. Gnanasekaran, Materials Research Forum (2012) (In press

5.28 N. Devictor (CEA, France)

In the framework of the French Act of the 28th of June 2006 about nuclear materials and waste management, a GENIV and actinides incineration demonstration prototype is to be commissioned in the 2020 decade.

This prototype, called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) sets out to demonstrate the progress made in SFR technology on an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability (see Abstract “Sodium-cooled Fast Reactors: the ASTRID plant project” by P. Le Coz, CEA).The paper resumes the R&D program led currently in support to the selection of ASTRID options, particularly on the following topics:

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- core design with the objective of improvement of its intrinsic behaviour in case of events with without scram, development of innovative third shutdown system, and of improvement of its monitoring (in-core neutronic monitoring, detection of local core damage, core geometry monitoring, or sodium boiling detection)

- improvement of the performances of decay heat removal system, and development of an efficient DHR system by structures,

- development of innovative fuel handling systems and energy conversion systems (especially gas ECS),

- development of a strategy in support to the limitation of core degradation consequences, including the R&D in support to the development of core catcher.

This R&D program takes into account the feedback experience of past reactors and SFR projects.

The paper will give information on the current development of experimental facilities for validation of innovative options and the new computer code generation, and on the schedule of the R&D in relationship to the ASTRID milestones.

5.29 K.K. Rajan (IGCAR, India)

Design, construction and operation of sodium cooled fast reactor, forms the second stage of the Indian Nuclear power programme. The PFBR is a 500MWe, sodium cooled, pool type, mixed oxide (MOX) fuelled fast reactor which is under an advanced stage of construction at Kalpakkam, India. The primary objective of the PFBR is to demonstrate techno-economic viability of fast breeder reactors on an industrial scale. Critical components of PFBR are needed to be tested in sodium at simulated reactor operating conditions and off normal conditions to qualify the design and ensure reliability. Testing of reactor components are also carried out in air and water depending on the nature of the components and the objective of testing. The ongoing PFBR component test program at FRTG, IGCAR includes testing of components such as in vessel and ex vessel fuel handling machines, viz. Transfer Arm and components of Inclined Fuel Transfer Machine (IFTM) - Primary Ramp and Primary Tilting Mechanism, Under Sodium Ultrasonic Scanner, Mutual inductance type Continuous level probes, DSRDM Electromagnet against thermal Shocks, Leak detector layout for PFBR and Gas Entrainment studies in sodium. This paper gives a brief report on the details of testing and qualification of the above PFBR components.

Key Words: PFBR, sodium technology, ultrasonic scanner, thermal shock, transfer arm, leak detectors, gas entrainment.

5.30 P. Selvaraj (IGCAR, India)

For future Fast Breeders Reactors (FBRs) it is decided to have a higher level of safety compared to PFBR (Prototype Fast Breeder Reactor). For future FBRs passive / additional safety features are to be introduced in reactor shutdown systems, so that the overall failure frequency of reactor protection system can be reduced to less than 10-7 per reactor year

(compared to 10-6 for PFBR). Decay heat removal after shutdown is one of the most important safety functions and must be accomplished with very high reliability. Safety criteria followed for plant design requires the non-availability of Decay Heat Removal function shall be less than 10-7 per reactor-year. Other than enhancing the safety the economy also has to be improved to make the future FBRs cost competitive. At IGCAR various R&D activities are directed to achieve the above objectives. The paper gives a brief overview of various R&D activities being carried out at IGCAR.

5.31 R.Sridharan aj (IGCAR, India)

High temperature energy conversion systems such as fast breeder reactors use liquid sodium as coolant to transfer heat from reactor core due to its favourable physical, chemical and nuclear properties [1]. The heat produced in the reactor core is transferred to steam generators by the sodium coolant circulating in the primary and secondary loops in order to produce steam required to generate power. It is essential to detect any event of leak of high pressure steam into secondary sodium coolant circuit at the steam generator using reliable sensors and initiate remedial action. This requires monitoring of hydrogen levels in secondary sodium in ppb levels as any micro leak of steam leak into sodium increases hydrogen concentration. The chemistry of sodium in the coolant circuits of the reactor is dominated by other non-metallic impurities like oxygen and carbon present in ppm levels in sodium. While pure liquid sodium is chemically compatible with the structural steels, presence of impurities like oxygen and carbon in sodium even in ppm levels can, lead to their corrosion and effect radioactivity transport from the reactor core to other areas [2, 3]. Hence, it is necessary to monitor these non metallic impurities in the coolant continuously using reliable sensors, to ensure safe reactor operation for its life time. Chemistry Group of Indira Gandhi Centre for Atomic Research has developed on-line electrochemical hydrogen meters [4] to monitor hydrogen levels in liquid sodium and in cover gas of secondary sodium circuit [5,6].

Solid electrolyte based Electrochemical hydrogen meters (ECHM) along with the instrumentation were calibrated in mini sodium loops and then installed in sodium circuits of plants at IGCAR such as steam generator test facility (SGTF), sodium-water reaction test facility (SOWART) and fast breeder test reactor (FBTR) and evaluated for their performance for several years. Based on the experience and performance of the sensors for plant related experiments, ECHMs are integrated in the upcoming prototype fast breeder reactor(PFBR). Under IGCAR-CEA collaboration studies on SFR Safety, an ECHM was installed on the intermediate circuit of PHENIX reactor, in order to assess its performances in realistic conditions. Experiments in these installations showed that the response of ECHM is in-phase with the diffusion based sensors for hydrogen detection and gave a satisfactory response to detect wider range of hydrogen levels in sodium. Experiments at Phenix showed that the response of ECHM to hydrogen injections is concordant with classical diffusion based sensors located at three different places in the plant.

To detect any steam leak in steam generator during start-up and low power operation of the reactor, diffusion based hydrogen detection system has been developed to measure hydrogen in the argon cover gas using a Thermal Conductivity Detector(TCD). The detection limit of this system is ~ 30 vppm of hydrogen. In order to extend the detection limit, a tin oxide based thin film hydrogen sensor has been developed and integrated with this system. This combined system shows a linear response to hydrogen from 2ppm onwards.

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Development of electrochemical sensors for monitoring oxygen and carbon impurities in sodium are the other significant efforts made in this laboratory [7, 8].

Current research activities and achievements on the performance of these sensors are exemplified in this presentation.

References:

[1] R.G. Palmer and A. Platt, in Fast Reactors, Temple Press Ltd., London (1961).

[2] H.U. Borgstedt and C.K. Mathews, Applied Chemistry of Alkali Metals, Plenum Press, New York (1987).

[3] M.R. Hobdell and C.A. Smith, J. Nucl. Mater., 110 (1982) 125. [4] Kitheri Joseph, K.Sujatha, S.Nagaraj, K.H.Mahendran, R. Sridharan, G.Periaswami,

T.Gnanasekaran, J. Nucl. Mater., 344 (2005) 285. [5] K.H. Mahendran, R. Sridharan, T. Gnanasekaran and G. Periaswami, Ind. Eng.

Chem. Res., 37 (1998) 1398. [6] E. Prabhu, V. Jayaraman, K.I. Gnanasekar, T. Gnanasekaran and G. Periaswami,

Asian J. Phys. 14 (2005) 33. [7] S. Rajan Babu, R. Ranganathan, S. Ramanamoorthy, K. Chandran, P. Muralidaran,

V. Ganesan, T. Gnanasekaran and G. Periaswami, Proceedings of the National Seminar on Physics and Technology of Sensors – 8 (2001) C-85.

[8] Rajesh Ganesan. V. Jayaraman, S. Rajan Babu, R. Sridharan and T. Gnanasekaran, J. Nucl. Science and Technol.,48, No.4 (2011) p.483-489.

5.32 P.Selvaraj (IGCAR, India)

For future Fast Breeders Reactors (FBRs) it is decided to have higher level of safety compared to PFBR (Prototype Fast Breeder Reactor). For future FBRs passive / additional safety features are to be introduced in reactor shutdown systems, so that the overall failure frequency of reactor protection system can be reduced to less than 10-7 per reactor year (compared to 10-6 for PFBR). Decay heat removal after shutdown is one of the most important safety functions and must be accomplished with very high reliability. Safety criteria followed for plant design requires the non-availability of Decay Heat Removal function shall be less than 10-7 per reactor-year. Other than enhancing the safety the economy also has to be improved to make the future FBRs cost competitive. At IGCAR various R&D activities are directed to achieve the above objectives. The paper gives a brief overview of various R&D activities being carried out at IGCAR.

5.33 R. Sridharan (IGCAR, India)

High temperature energy conversion systems such as fast breeder reactors use liquid sodium as coolant to transfer heat from reactor core due to its favourable physical, chemical and nuclear properties [1]. The heat produced in the reactor core is transferred to steam generators by the sodium coolant circulating in the primary and secondary loops in order to produce steam required to generate power. It is essential to detect any event of leak of high pressure steam into secondary sodium coolant circuit at the steam generator using reliable sensors and initiate remedial action. This requires monitoring of hydrogen levels in secondary sodium in ppb levels as any micro leak of steam leak into sodium increases hydrogen concentration. The

chemistry of sodium in the coolant circuits of the reactor is dominated by other non-metallic impurities like oxygen and carbon present in ppm levels in sodium. While pure liquid sodium is chemically compatible with the structural steels, presence of impurities like oxygen and carbon in sodium even in ppm levels can, lead to their corrosion and effect radioactivity transport from the reactor core to other areas [2, 3]. Hence, it is necessary to monitor these non metallic impurities in the coolant continuously using reliable sensors, to ensure safe reactor operation for its life time. Chemistry Group of Indira Gandhi Centre for Atomic Research has developed on-line electrochemical hydrogen meters [4] to monitor hydrogen levels in liquid sodium and in cover gas of secondary sodium circuit [5,6].

Solid electrolyte based Electrochemical hydrogen meters (ECHM) along with the instrumentation were calibrated in mini sodium loops and then installed in sodium circuits of plants at IGCAR such as steam generator test facility (SGTF), sodium-water reaction test facility (SOWART) and fast breeder test reactor (FBTR) and evaluated for their performance for several years. Based on the experience and performance of the sensors for plant related experiments, ECHMs are integrated in the upcoming prototype fast breeder reactor (PFBR). Under IGCAR-CEA collaboration studies on SFR Safety, an ECHM was installed on the intermediate circuit of PHENIX reactor, in order to assess its performances in realistic conditions. Experiments in these installations showed that the response of ECHM is in-phase with the diffusion based sensors for hydrogen detection and gave a satisfactory response to detect wider range of hydrogen levels in sodium. Experiments at Phenix showed that the response of ECHM to hydrogen injections is concordant with classical diffusion based sensors located at three different places in the plant.

To detect any steam leak in steam generator during start-up and low power operation of the reactor, diffusion based hydrogen detection system has been developed to measure hydrogen in the argon cover gas using a Thermal Conductivity Detector(TCD). The detection limit of this system is ~ 30 vppm of hydrogen. In order to extend the detection limit, a tin oxide based thin film hydrogen sensor has been developed and integrated with this system. This combined system shows a linear response to hydrogen from 2ppm onwards.

Development of electrochemical sensors for monitoring oxygen and carbon impurities in sodium are the other significant efforts made in this laboratory [7, 8].

Current research activities and achievements on the performance of these sensors are exemplified in this presentation.

References:

[1] R.G. Palmer and A. Platt, in Fast Reactors, Temple Press Ltd., London (1961). [2] H.U. Borgstedt and C.K. Mathews, Applied Chemistry of Alkali Metals, Plenum

Press, New York (1987). [3] M.R. Hobdell and C.A. Smith, J. Nucl. Mater., 110 (1982) 125. [4] Kitheri Joseph, K.Sujatha, S.Nagaraj, K.H.Mahendran, R. Sridharan, G.Periaswami,

T.Gnanasekaran, J. Nucl. Mater., 344 (2005) 285. [5] K.H. Mahendran, R. Sridharan, T. Gnanasekaran and G. Periaswami, Ind. Eng.

Chem. Res., 37 (1998) 1398. [6] E. Prabhu, V. Jayaraman, K.I. Gnanasekar, T. Gnanasekaran and G. Periaswami,

Asian J. Phys. 14 (2005) 33. [7] S. Rajan Babu, R. Ranganathan, S. Ramanamoorthy, K. Chandran, P. Muralidaran,

V. Ganesan, T. Gnanasekaran and G. Periaswami, Proceedings of the National

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Seminar on Physics and Technology of Sensors – 8 (2001) C-85. [8] Rajesh Ganesan. V. Jayaraman, S. Rajan Babu, R. Sridharan and T. Gnanasekaran,

J. Nucl. Science and Technol.,48, No.4 (2011) p.483-489.

5.34 V.A. Suresh Kumar (IGCAR, India)

India’s nuclear power programme has been structured in three stages. The second stage started with the construction, commissioning and operation of Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Operating experience and better understanding of the technologies involved in FBTR has enabled us to design a 500 MWe Prototype Fast Breeder Reactor (PFBR). Construction of the PFBR is in the advanced stage at Kalpakkam. Steam Generator (SG) is one of the most important components in sodium cooled fast reactor and based on the available operating experience of various Liquid Metal Fast Breeder Reactors (LMFBRs), it can be considered as a major component that decides plant availability. Steam generators with lesser quantity of water are preferred in FBRs from the consideration of sodium-water reaction. Hence, once through steam generators (OTSG) are employed in FBRs. A straight tube SG with expansion bend is selected for PFBR. A 5.7 MWt capacity Steam Generator Test Facility (SGTF) has been set up in IGCAR, Kalpakkam to test a 19-tube model steam generator of PFBR. The material of construction, tube dimensions and operating parameters of SGTF SG are similar to that of PFBR SG except the number of tubes. Many experiments were successfully conducted on SGTF SG and the results have provided valuable feed back to the design of SG of PFBR and future FBRs. This paper brings out various experiments conducted on the model steam generator.

Key words: Fast breeder reactor, once through steam generator, experiements

5.35 B.K. Sreedhar (IGCAR, India)

Vertical centrifugal pumps are used for circulation of liquid sodium coolant in the primary and secondary circuits of fast reactors. These pumps employ oil lubricated bearings and seals to support the pump rotor above the cover gas space and to isolate the cover gas from the atmosphere. Although engineered safety features are built into the system and adequate operational procedures are employed to prevent any oil ingress into sodium, any incident involving inadvertent contamination of sodium with oil can lead to long and costly shutdown. It is therefore prudent to consider oil free designs for bearings and seals for pumps of future reactors. A design variant that enables elimination of oil from bearings and seals involves use of Active Magnetic Bearings (AMB) and ferrofluid seals. AMB are contactless bearings that support the rotor assembly through magnetic forces generated by electromagnets. Ferrofluid seals employ a liquid that responds to an external magnetic field and functions as a 'liquid O ring' to achieve sealing. This paper discusses the work in progress towards development of Active Magnetic Bearing (AMB) and ferrofluid seal for centrifugal sodium pump.

5.36 V. Prakash (IGCAR, India)

Sodium cooled Fast Breeder Reactors form the bastion of Indian nuclear programme in the years to come. Prototype Fast Breeder Reactor (PFBR) is 500MWe sodium cooled pool type reactor, which is under construction at Kalpakkam, India. Sodium flow measurement is of prime importance both from the operational and safety point of view of a fast reactor. Primary sodium pump discharge, secondary sodium circuit, safety grade decay heat removal circuit and auxiliary sodium circuits are the critical locations in an FBR, where sodium flow measurement is of importance. Apart from this, sodium flow measurement in subassemblies in the core facilitates the detection of flow blockage in the subassemblies.

Permanent magnet flow meter is widely used for pipe flow measurement in sodium systems, however it cannot be considered for applications like subassembly and pump discharge flow measurements. From the safety point of view flow through the core should be assured under all operating conditions in an FBR. This poses the challenge of developing a flow sensor which can withstand the high temperature sodium environment and can meet the dimensional constraints and be amenable to maintenance. An eddy current flow meter is designed and developed for sodium flow measurement in PFBR and the developed probe is tested for its performance in a sodium test rig. The developed ECFM is also used for core flow measurements in FBTR, India and in Phenix reactor,France, during its end of life test before decommissioning. This paper discusses the activities related to the development, calibration, testing of ECFM probe for PFBR and results of its utilization for flow measurement in FBTR and in Phenix reactor, France.

5.37 F. Roelofs (NRG, The Netherlands)

Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics.

The heart of every nuclear reactor is the core where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to a coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential.

The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterisation of the liquid metal flow structure in the above core region is very important.

This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction required between experiments and advanced simulations. Furthermore, it will highlight the latest developments using Computational Fluid Dynamics (CFD) simulation techniques worldwide with a special focus on the developments and achievements within NRG in the Netherlands.

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With respect to fuel assembly thermal hydraulics, the latest developments on simulation of fuel assemblies with wire wraps will be highlighted. As well defined experimental data is hard and/or expensive to obtain, detailed CFD with advanced turbulence modeling and large computational resources is used to create reliable reference data. Furthermore, simulations applying various turbulence models and different codes are inter-compared to gain confidence in the numerical results.

With respect to pool thermal hydraulics, the latest developments using CFD with state-of-the-art numerical grid construction and turbulence modeling will be demonstrated. Although experimental data from water experiments is available in this case for specific designs, a proper validation for the CFD simulations is hard to achieve. Again, simulations using different numerical codes, grids, and turbulence models are inter-compared to gain confidence in the numerical results.

5.38 R.Vijayashree (IGCAR, India)

The environmental pollution caused by the release of green house gases by fossil power plants, associated global warming and also depletion of all known sources of fossil fuels are the factors that are driving the world to look for alternate sources of power generation. For the nuclear reactors to gain their legitimate role as a clean source of power, the possibility of major accidents (like that of Three Mile Island in 1979, Chernobyl in 1986 and Fukushima in 2011) is one of the major obstacles. Hence, the augmentation of nuclear reactor safety is very essential for any reactor to gain public acceptance and global importance. The shutdown system and decay heat removal system are the two important subsystems that ensure the nuclear reactor safety. The reliability of the shutdown system can be considerably enhanced by using diverse and passive principles for shutting down the reactor.

Any shutdown system comprises of reactor protection system and Actuation system. The sensors signal processing units and logic are under the reactor protection system. The absorber rods and their drive mechanism form the actuation system. .The reliability analysis of the existing configuration of the shutdown system of PFBR has indicated that they are contributing almost equally to the failure of shutdown system.

Passive means of triggering the introduction of neutron poison in the system on global faults helps to enhance the reliability by augmenting to that of reactor protection system. However this alone is not sufficient to enhance reliability because the unreliability due to actuation system starts dominating. Therefore it is proposed to introduce a third shutdown system based on liquid poison/absorber balls having both active and passive triggers for the introduction of poison into the core on global faults In addition the existing design of electromagnet of Diverse safety rod mechanism is enhanced to that of temperature sensitive Electromagnet. Stroke limiting devise which is an engineered safety feature is also proposed to be introduced in the existing design of Control and Safety Rod Drive Mechanism. This paper presents a brief description of the above mentioned enhancement proposed in the existing design of shut down system. The conceptual design of third shutdown system under consideration is also presented.

5.39 S.R. Choi (KAERI, Republic of Korea)

In the SFR core thermal hydraulic design, it is a critical issue to quantify the safety margins including uncertainties of design factors and limits. The deviations from the nominal values need to be quantitatively considered by statistical thermal design methods. The hot channel factors (HCF) were employed to evaluate the uncertainty in the early design such as the CRBRP. The improved thermal design procedure (ITDP) calculates the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. Another way to consider the uncertainties is to use the Monte Carlo method (MCM). In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. It is able to directly estimate the uncertainty effects and propagation characteristics for the present thermal-hydraulic model. However, it requires a huge computation time to get a reliable result because the accuracy is dependent on the sampling size. This paper describes comparative studies of both thermal hydraulic models and statistical analysis methods performed during the development of a Sodium-Cooled Fast Reactor. In order to evaluate the eutectic limit between fuel slug and cladding, the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code is modified to estimate the uncertainties of cladding inner wall temperatures to check the eutectic limit temperature. A subchannel analysis and hot channel factors-based uncertainty evaluation system is established to estimate the core Thermofluidic uncertainties using the MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors) code and the results are compared to those of the SLTHEN code. In addition, the MATRA-LMR code is connected to the improved thermal design procedure and Monte Carlo method to analyze effect of the statistical techniques. The calculated results of the hot channel factor were close to those of the improved thermal design procedure owing to their similar statistical analyses. In the present analysis, it was clear that the dominant input parameter was the fuel thermal conductivity considering its uncertainty value and temperature increase. Statistical uncertainty was also evaluated by incorporating the Monte Carlo sampling into the thermal-hydraulic analysis code MATRA-LMR. About 30,000 calculations were conducted to eliminate the sampling size effect and verify the present uncertainty assessing method. The average and standard deviation were slightly larger than those of the other methods. 5.40 S. Bortot (Politecnico di Milano, Italy)

The Lead-cooled Fast Reactor (LFR) is under development worldwide as a very promising fast neutron system to be operated in a closed fuel cycle. The LFR complete development requires - as a fundamental intermediate step - the realization of a demonstration plant intended to prove the viability of technology as well as the overall system behavior. At a technical level, development of a LFR demonstrator is to be realized through the LEADER (Lead-cooled European Advanced Demonstration Reactor) project of the European Commission’s 7th Framework Programme, which is strongly committed to the conceptual design of a low-cost scaled-down reactor fully representative of the industrial size system named ALFRED (Advanced Lead Fast Reactor European Demonstrator). Due to the need of investigating reactor responses to power and temperature transients, both a lumped-parameter and an object-oriented one-dimensional approach has been undertaken to provide helpful tools in this early phase of the preconceptual design allowing a relatively quick, qualitative analysis of fundamental stability and dynamics aspects that cannot be left aside when refining or even finalizing the system configuration. As far as the former are concerned, investigations

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have been focused on a preliminary core proposal and the impact of fuel burn-up and of fundamental parameters on the system stability has been evaluated considering both a stand-alone core and a primary loop configuration. As a major result of this analysis, a crucial role played by the coolant density coefficient has been outlined, with subsequent important feedback for the reactor designer. Furthermore, the object-oriented model has been employed to perform transient design-basis analyses by simulating typical scenarios such as Unprotected Loss of Heat Sink (ULOHS), Transient of Over Power (UTOP) and Loss Of Flow (ULOF) events.

5.41 K.Velusamy (IGCAR, India)

Thermal hydraulics plays an important role in fast breeder reactor (FBR) design. Liquid sodium is the choice for core cooling due to its large heat transfer coefficient and high boiling point, to have a compact reactor working at low pressures. Unfortunately, large heat transfer coefficient of sodium leads to large transient temperature variations in the structures, during plant transients, with minimum attenuation in the boundary layer. The coefficient of volumetric expansion of sodium and the possible sodium temperature variations are very large, which are conveniently utilized for natural convection cooling of the core. But, coexistence of large volumes of hot and cold sodium pools in the same vessel leads to special phenomena such as thermal stratification and free level fluctuations, and the associated temperature fluctuations in the structures. Free level fluctuations can lead to argon gas entrainment in the sodium, which eventually has the potential to lead to reactivity perturbations. High operating temperature and large thermal stress control the structural design. This demands accurate knowledge of temperature distribution in the structures, which has to be obtained through multi-mode heat transfer analyses. Since the components are thin and slender, accurate knowledge of velocity distribution in the components is essential to design them against flow induced vibration. Also, the components have to be optimized to provide acceptable flow and pressure distributions and thus, single / multi-dimensional thermal hydraulics assumes special significance in FBR design. Difficulties encountered in performing large-scale sodium experiments and inability of water to simulate thermal aspects of sodium repose heavy responsibility on thermal hydraulic analyst.

Safety of the plant during various categories of Design Basis Events (DBE) is evaluated by a one-dimensional in-house developed plant dynamics code DYANA-P, which has models for core, hot / cold pools, intermediate heat exchangers, sodium pumps, steam generator, primary / secondary sodium circuits, neutron kinetics and reactivity feedback. Safety analysis of the plant during safety grade decay removal conditions is carried out using another in-house developed code DHDYN. These codes are validated against test data obtained from Fast Breeder Test Reactor. By using these codes all the identified DBE have been analyzed and the safety has been ensured for the Prototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction. For the purpose of multi-dimensional Computational Fluid Dynamic (CFD) investigations, in-house CFD codes THYC-2D / THYC-3D along with commercial codes viz., PHOENICS, STAR-CD, FLUIDYN and CFD-Expert are employed. Predictions of these codes have been validated against in-house sodium / water experiments performed. Adopting these codes judiciously, a sound thermal hydraulic design has been arrived at for various sub-systems PFBR. Despite the capabilitites established at the centre, continuous interacton is maintained with premier academic institutions and national / international bodies, through collaborative research projects, to enhance the quantum and

quality of output. The full paper will give details of the thermal hydraulic studies performed for PFBR and the future directions planned in the domain of thermal hydraulic research.

5.42 A.P.Sorokin (IPPE, Russia)

The substantiation of thermohydraulic characteristics and fast reactors safety include of studying of a wide range of problems for stationary regimes, the transitive regimes controlled by protection NPP, and the accident regimes of NPP connected with failure of protection system, destruction of elements of a circulation contour and also hypothetical, improbable accidents.

Experimental investigations are spent, as a rule, for studying of thermohydraulic processes features in the reactor vessel and units of reactor installation and receptions of initial data for testing and verification of numerical programs. For this purpose in Rosatom there is considerable experimental base concentrated mainly in SSC RF-IPPE.

Recently the methods and codes for mathematical modelling of heat and mass transfer in the reactor vessel and units of reactor installation have gained the considerable development. Thermohydraulic calculations lead to raise problems about joint heat exchange in system a liquid – design element that is to the solution of the interfaced problems of heat exchange.

With reference to fast reactors with the sodium coolant the investigations on thermohydraulic substantiation of fast reactors are spent for a wide range of problems:

− the thermohydraulic of reactor core and individual fuel assemblies; − the processes of mixed and natural convection of coolant in the reactor vessel and in

separate fuel assemblies; − the processes in the upper mixture chamber including interaction of jets over fuel

assembly heads, temperature pulsations in coolant and in construction units; − the processes of capture of the gas from a surface of sodium in a gas cavity of a reactor

and its carrying along on a contour of installation; − the thermohydraulic characteristics of fuel assembly with blockage of a part of cross

section; − the processes of heat and mass transfer connected with formation, transport,

sedimentation and accumulation of impurity in contours with sodium and in protective gas of fast reactors;

− hydrodynamics and heat exchange processes in steam genetators of fast reactors in various regimes of their work;

− the vibration strength of members of fast reactors; − the processes of arising and development of boiling, including regimes of decay heat

removal with a natural circulation of coolant in fast reactors; − the interaction of sodium with water at destruction of tubes of steam generators by type

"sodium-water"; − the sodium fires at a depressurization of circuits with formation of aerosols; − the interaction of corium with sodium at fuel fusion in development of heavy accident

including arising of steam explosion.

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Many of the listed problems are complex and demand cooperation studying of thermohydraulic, heat and mass exchange, neutrono-physical and thermomechanical processes.

Thermophysical investigations are baseline for development of methods and equipments for diagnostic and control of parameters of fast reactors and development of means of an accident protection including a passive safety system.

I focus attention on separate problem questions of thermohydraulic substantiations of characteristics and fast reactors safety with the sodium coolant which studying represents now a special urgency. These are questions:

− the natural convection developments in a contour of circulation and individual fuel assemblies;

− the processes of arising and development of boiling, including regimes of decay heat removal with a natural circulation of coolant in fast reactors;

− the interaction of corium with sodium; − the heat exchange and stability of circulation in steam generators of fast reactors; − the modelling of mass transfer of impurity in contours of reactor installations; − the development of verification tests for thermohydraulic codes; − the development and verification of numerical codes for modelling of thermohydraulic

in reactor vessel and construction units of fast reactors including transitive and accidental processes in contours and equipment elements.

With reference to fast reactors with cooling by heavy liquid metal coolants are represented an urgency of investigations of such problem questions, as:

− the heat exchange and temperature regimes of fuel rods in a reactor core of fast reactors; − the heat exchange in tube bundles of steam generator; − the interaction of corium with heavy metal coolants; − the interaction of lead (alloys on the basis of lead) with water at leakage of the steam

generator.

5.43 K. Tucek (IET, JRC, European Commission)

The Institute for Energy and Transport (IET) is one of the seven scientific institutes of the Joint Research Centre of the European Commission. Its mission is to provide support to definition, implementation, and monitoring of European Union policies related to both nuclear and non-nuclear energy, and to transport. The IET is based both in Petten, the Netherlands, and Ispra, Italy, and has a multidisciplinary team of around 300 academic, technical, and support staff. The research areas include, among others, sustainable transport, renewable energies incl. solar, photovoltaics and biomass, sustainable & safe nuclear energy for present & future reactor system, pre-normative materials testing, energy techno-economic assessment, hydrogen and fuel cells, energy efficiency, and security of energy supply.

To support implementation of the European Strategic Energy Technology Plan (SET-plan) and the Industrial Initiatives of the Sustainable Nuclear Energy Technology Platform (ESNII and NC2I, under construction) the IET works on performance assessment, design optimization, and safety evaluation of various next generation nuclear fission (Generation IV)

energy systems. In collaboration with the Generation IV International Forum (GIF) and with European Technical Support Organizations, IET participates in development of an integrated safety assessment methodology for Generation-IV reactor systems aiming at establishment of a widely applicable, technology-neutral concept. Mostly as part of European collaborative projects, IET is also engaged in safety-relevant experimental research on advanced fuels, components and materials, in simulation as well as modelling to support experimental research and to assess performance of Generation-IV systems. This includes irradiation testing of fuels for fast reactors and high temperature reactors, of transmutation fuels and candidate structural materials for several advanced reactor concepts. Out-of-pile testing and qualification methods in corrosive environments up to very high temperatures are also being developed and applied. This work contributes to several European collaborative projects, to certain GIF projects, to the Joint Research Program on Nuclear Materials of the European Energy Research Alliance, and to related IAEA and OECD/NEA activities.

5.44 K.Velusamy (IGCAR, India)

Thermal hydraulics plays an important role in fast breeder reactor (FBR) design. Liquid sodium is the choice for core cooling due to its large heat transfer coefficient and high boiling point, to have a compact reactor working at low pressures. Unfortunately, large heat transfer coefficient of sodium leads to large transient temperature variations in the structures, during plant transients, with minimum attenuation in the boundary layer. The coefficient of volumetric expansion of sodium and the possible sodium temperature variations are very large, which are conveniently utilized for natural convection cooling of the core. However, coexistence of large volumes of hot and cold sodium pools in the same vessel leads to a special phenomena such as thermal stratification and free level fluctuations, and the associated temperature fluctuations in the structures. Free level fluctuations can lead to argon gas entrainment in the sodium, which eventually has the potential to lead to reactivity perturbations. High operating temperature and large thermal stress control the structural design. This demands accurate knowledge of temperature distribution in the structures, which has to be obtained through multi-mode heat transfer analyses. Since the components are thin and slender, accurate knowledge of velocity distribution in the components is essential to design them against flow induced vibration. Also, the components have to be optimized to provide acceptable flow and pressure distributions and thus, single / multi-dimensional thermal hydraulics assumes special significance in FBR design. Difficulties encountered in performing large-scale sodium experiments and inability of water to simulate thermal aspects of sodium repose heavy responsibility on thermal hydraulic analyst.

Safety of the plant during various categories of Design Basis Events (DBE) is evaluated by a one-dimensional in-house developed plant dynamics code DYANA-P, which has models for core, hot / cold pools, intermediate heat exchangers, sodium pumps, steam generator, primary / secondary sodium circuits, neutron kinetics and reactivity feedback. Safety analysis of the plant during safety grade decay removal conditions is carried out using another in-house developed code DHDYN. These codes are validated against test data obtained from Fast Breeder Test Reactor. By using these codes all the identified DBE have been analyzed and the safety has been ensured for the Prototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction. For the purpose of multi-dimensional Computational Fluid Dynamic (CFD) investigations, in-house CFD codes THYC-2D / THYC-3D along with commercial codes viz., PHOENICS, STAR-CD, FLUIDYN and CFD-Expert are employed. Predictions of these codes have been validated against in-house sodium / water experiments

48

performed. Adopting these codes judiciously, a sound thermal hydraulic design has been arrived at for various sub-systems PFBR. Despite the capabilitites established at the centre, continuous interacton is maintained with premier academic institutions and national / international bodies, through collaborative research projects, to enhance the quantum and quality of output. The full paper will give details of the thermal hydraulic studies performed for PFBR and the future directions planned in the domain of thermal hydraulic research.

5.45 V.Vinod (IGCAR, India)

In the present scenario the decay heat removal from the nuclear reactors is of great importance. In the sodium cooled pool type fast breeder reactors, decay heat generated from the reactor core followed by an unprotected reactor trip is transferred to the sodium pool initially by forced circulation and then by the buoyancy driven flow. The heat is removed from the sodium pool by means of dedicated passive decay heat removal systems. To demonstrate and validate the different decay heat removal mechanisms in the prototype fast breeder reactor (PFBR) different studies have been conducted in sodium and water medium. The heat removal mechanism from the core to pool and then to the decay heat exchanger was studied in a 1/4th scaled water model of the reactor. The effect of inter wrapper flow on decay heat removal was studied separately with a full scale slab model of reactor core in water. Further the decay heat removal from the sodium hot pool to the atmosphere through the sodium to sodium decay heat exchanger and sodium to air heat exchanger was studied with a scaled down sodium model of the system in a facility called SADHANA. The methodology of the experimental study on decay heat removal for PFBR and the results are discussed.

5.47 E. Hemanth Rao (IGCAR, India)

Choice of sodium as coolant in fast reactors is being challenged despite the vast operating experience accumulated worldwide over 400 reactor years, due to few sodium leak incidents in earlier reactors Monju, PFR, Phenix, Superphenix etc. However, fairly good experience with sodium systems and development in technology drives majority of fast reactor programs continue with sodium. Moreover, Sodium cooled Fast Reactors (SFRs) are considered to be robust and ready for commercialization at the earliest among other types of fast reactors. In this context, it is necessary to investigate a few safety issues such ase consequences sodium fire, development of techniques for effective extinguishment and mitigation. Development of suitable codes for prediction of burning rates during different sodium fire scenarios under realistic situations, pressure development and consequent aerosol distribution would comprehensively raise the confidence in the operation of future SFRs. Many studies were carried out in the past at IGCAR, India towards design validation of engineered safety features, which are adopted for Prototype Fast Breeder Reactor (PFBR). In continuation, an extensive research program has been taken up which includes experiments from scientific to engineering scale with a motive to generate vital data on sodium fires and related safety issues. Few facilities are setup for aiding the experimental programs and suitable numerical codes are being developed. Attention is paid on consequences of sodium fires, secondary sodium fires and development of new, effective extinguishing powders as well as sodium resistant concrete. This paper outlines few important results, ongoing studies and future road map for research program in the field of sodium fire safety to support ambitious future SFR program.

5.48 S. Kumar Das (IGCAR, India)

Post accident heat removal (PAHR) in Fast Breeder Reactors (FBR) carries lot of importance in bringing the reactor to safe shutdown configuration and ensuring public safety. Design and implementation of effective safety features, demand thorough understanding of the critical phenomena involved in sequences during sever accident. Key driving factors for molten fuel jet breakup, interaction with coolant, fragmentation leading to generation of debris and relocation / bed forming characteristics, require in-depth investigation for accurate prediction of the consequences. Morphological characteristics of accumulated debris and available flow paths for coolant in bottom plenum, play important role in achieving non critical and effectively coolable geometry for the relocated fuel. Combined effort for numerical and experimental studies has been taken up at Indira Gandhi Centre for Atomic Research (IGCAR) under severe accident research program, with primary objective to address the PAHR issues highlighted above. Series of dedicated experimental facilities have been setup in stages and preliminary trials are being conducted to generate the database for development and validation of numerical models. Few of the interesting experimental results and comparisons of data with numerical findings are discussed.

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6. RECOMMENDATIONS FOR REAL TIME RESEARCH AREAS

6.1 Research Areas in Fast reactor physics and technology

This meeting has identified several areas where further international effort to be focused. It has been agreed by all representatives of the Member States that sharing of the experiences, best practice as well as methodologies should be supported by the IAEA. In view of effective networking on Fast reactor Physics and Technology several recommendation s have been formulated and agree upon the closing session of the meeting. Experts agreed that postulated need should be implemented through following IAEA activities.

(i) Agency should organise meeting of this series on Fast reactor physics and technology periodically (once in 2 years) in a country having interest in FBR technology.

(ii) Development of design rules for fast reactor technologies in respect of passive safety systems. IAEA should take a lead role is safety criteria for FBRs and codes and guides specific for fast reactors to start with for the sodium and other liquid metals cooled fast reactors in particular on fire incidents, severe accidents and in-service inspection.

(iii) IAEA meeting on impact of Fukishima accident on the safety aspects of existing and planned FBRs in particular approach towards beyond design basis events and margins on design basis flood and safe shut down earthquake parameters. Further adaptation of LWR safety criteria to be considered for further application in FBR deployment and operation.

(iv) A technical seminar under IAEA umbrella on experimental facilities planned in Member States towards the fast reactors with its objectives and norms for sharing of results would be worth considering. This includes code validation for safety analysis, specifically under accident and transient conditions, analysis of severe accidents with core melting, fuel testing and power ramping experiments, qualification procedures for assessment of long-term stability of critical components (creep, fatigue, swelling etc.).

(v) Agency should prepare the training on Fast reactor coolant technologies and basic principle of passive safety and design advantages and operational experience. This also includes the support of international conferences through available mechanisms of Technical Cooperation.

7. ANNEXES

7.1 Agenda of the meeting

Day 1 : Nov 14 , 2011

Program / Topic Speaker TIME

Session I : Inauguration

Welcome address

Inaugural address

Inaugural Address

Feedback from IAEA TWG-FR

Dr. P. Chellapandi, Dir-NSEG

Shri. S.C. Chetal, Dir-IGCAR

A. Zeman, IAEA/NAPC

S. Monti, IAEA/NENP

9.30-9.40

9.40-9.55

9.55-10.05

10.05-10.15

High Tea 10.15-10.45

Session II : Status of National Programmes (Chair: S.Monti, IAEA/NENP)

Presentation by Belgium D. De Bruyn 10.45-11.00

Presentation by China K. Zhou 11.00-11.15

Presentation by Czech Republic M. Brumovsky 11.15-11.30

Presentation by European Commission K. Tucek 11.30-11.45

Presentation by France François GAUCHE 11.45-12.00

Presentation by Germany B. Merk 12.00-12.15

Presentation by India P. Chellapandi 12.15-12.30

Presentation by Italy A. Gessi 12.30-12.45

Presentation by Japan T.Asayama 12.45-13.00

Lunch 13.00-14.00

Presentation by Korea S. Choi 14.00-14.15

Presentation by Netherlands F. Roelofs 14.15-14.30

Presentation by Russian Federation O.Saraev/V.Lemenkov 14.30-14.45

Presentation by Sweden J. Wallenius 14.45-15.00

Presentation by Ukraine S. Fomin 15.00-15.15

Presentation by USA TBC (USA) 15.15-15.30

Discussion on recent issues regarding FR All participants 15.30-15.45

Tea 15.45-16.00

Session III : Reactor Physics and Core Design (Chair: K. Tucek)

Physical Basis of Advanced fast reactor working S. Fomin (Ukraine) 16.00-16.30

52

in Nuclear Burning Wave Regime

Axial and Radial shield optimization for future FBR with Ferroboron

D. Sunil Kumar (India) 16.30-17.00

A Moderation layer to improve the safety behaviour of sodium cooled fast reactors

B. Merk (Germany) 17.00-17.30

Detection of coolant void and fuel assembly dislocations in fast reactors using changes in the neutron spectrum

P. Wolniewicz (Sweden) 17.30-18.00

Day 2 : Nov 15, 2011

Session III : Reactor Physics and Core Design (con’t)

Development and testing for fast reactor fuels at the HFR Materials test reactor

F. Klassen (Netherlands) 09.30-10.00

Inherent safety aspects of Metal fuelled FBRs T. Satyasheela (India) 10.00-10.30

Physics aspects of Metal fuelled fast reactors D. K. Mohapatra (India) 10.30-11.00

Tea 11.00-11.15

Comparison of measured and calculated Keff values during FBTR operation

G. Raghukumar (India) 11.15-11.45

Summary by Chariman 11.45-12.00

Session IV : Advanced Reactor Designs including ADS (Chair: S. Monti)

Fast Reactor Programme in France F. Gauche’ (France) 12.00-12.30

Sodium cooled fast reactors : The ASTRID Plant Project

P.Le Coz (France) 12.30-13.00

MYRRHA, the Belgian flexible fast spectrum irradiation facility

D. De Bruyn (Belgium) 13.00-13.30

BREST OD 30 Design and development issues

V. Lemekhov (Russia) 13.30-14.00

Lunch 14.00-14.30

Design approach to improve economy for future FBRs

P. Chellapandi (India) 14.30-15.00

Reactor core P. Puthiyavinayagam (India)

15.00-15.15

Reactor Assembly V. Rajan Babu (India) 15.15-15.30

Shutdown systems R. Vijayashree (India) 15.30-15.45

Tea 15.45-16.00

Fuel handling systems S. Raghupathy (India) 16.00-16.15

Heat transport systems V. Balasubramaniyan (India) 16.15-16.30

Instrumentation & control K. Madhusoodhanan (India) 16.30-16.45

Physics studies on thorium utilisation and sub-criticality measurement in Accelerator Driven Systems

S.B.Degweker (India) 16.45-17.15

Discussion and Summary by Chairman 17.15-17.30

Day 3 : Nov 16, 2011

Technical visit to IGCAR facilities and PFBR Interested participants 09.30-13.00

Lunch 13.00-14.00

Session V : Structural Materials Development (Chair: A. Zeman)

Structural materials issues in GFR M. Brumovsky (Czech Republic)

14.00-14.30

Materials development for fast reactor applications

T. Jayakumar (India) 14.30-15.00

Assessment of advanced core structural materials for radiation damage using ion beams and materials modeling

B.K. Panigrahi (India) 15.00-15.30

Mechanical property evaluation of irradiated FBTR core structural materials

V. Karthik (India) 15.30-16.00

Structural materials for Japanese FBR T.Asayama (Japan) 16.00-16.30

Summary by Chairman

Tea 16.30-16.45

Session VI : Coolant Technology and Component Development (Chair: C. Latge’)

Liquid metal fast reactors cooled by Na or Pb : Respective issues and synergies

C.Latge (France) 16.45-17.15

Pb and LBE : a Technological comparison A. Gessi (Italy) 17.15-17.45

Day 4 : Nov 17, 2011

Session VI : Coolant Technology and Component Development (con’t)

54

ELECTRA : European lead cooled training reactor

J.Wallenius (Sweden) 09.30-10.00

Studies on chemical compatibility in lead-steel systems and development of oxygen monitors for lead and LBE systems

Rajesh Ganesan (India) 10.00-10.30

R&D program in support to ASTRID project N. Devictor (France) 10.30-11.00

Tea 11.00-11.30

Testing and Qualification of PFBR components in air and sodium

K.K. Rajan (India) 11.30-12.00

R&D activities for future FBRs P. Selvaraj (India) 12.00-12.30

Sensors for on-line monitoring of non metallic impurities in liquid sodium

R. Sridharan (India) 12.30-13.00

Lunch 13.00-14.00

Performance evaluation of PFBR model steam generator

V.A. Suresh Kumar (India) 14.00-14.30

Development of magnetic bearings and ferro fluid seals towards oil free sodium pumps

B.K. Sreedhar (India) 14.30-15.00

1 Utilisation of eddy current flow meter for sodium flow measurements in FBRs

V. Prakash (India) 15.00-15.30

Tea 15.30-16.00

2 Testing and qualification of PFBR shutdown systems

R. Vijayashree (India) 16.00-16.30

Summary by Chairman 16.30-17.00

Day 5 : Nov 18, 2011

Session VII : Thermal Hydraulics (chair TBD from Russia)

Fuel assembly and Pool thermal hydraulics for fast reactors

F. Roelofs (Netherlands) 09.30-10.00

Uncertainty assessment of core thermal hydraulic analysis for a sodium cooled fast reactor

C. Sun-rock (Korea) 10.00-10.30

Dynamics and stability analyses of the European LFR demonstrator ALFRED

S. Bortot (Italy) 10.30-11.00

Tea 11.00-11.15

Synthesis of Thermal hydraulic studies for pool type FBRs

K. Velusamy (India) 11.15-11.45

Actual problems of fast reactors thermohydraulics A.P. Sorokin (IPPE) 11.45-12.15

Summary by Chairman 12.15-12.30

Session VIII : Reactor Safety (Chair F. Gauche’)

Generation IV Reactor Safety and Material Research by the Institute for Energy and Transport at the European Commission’s Joint Research Centre

K. Tucek (European Commission)

12.30-13.00

Experiments for demonstrating passive SGDHR system of PFBR

G. Padmakumar (India) 13.00-13.30

Lunch 13.30-14.00

Innovative safety features of future FBRs P. Chellapandi (India) 14.00-14.30

Numerical and Experimental studies related to sodium fire

E. Hemanth Rao (India) 14.30-15.00

Post Accident Heat Removal : Numerical and Experimental simulation

S. Kumar Das (India) 15.00-15.30

Summary by Chairman 15.30-15.45

Tea 15.45-16.00

Session IX : Conclusion (Chair: S. C. Chetal)

Summary & Conclusions, proposals for future collaboration/activities and minutes of meeting

All Participants 16.00-17.30

Note : Time of presentation includes discussion

56

7.2 List of participants

No Name of participant Country Email address

1 Didier De Bruyn Belgium didier.de.bruyn@sckcen.be

2 Keyuan Zhou China keyuan@ciae.ac.cn

3 Milan Brumovsky Czech Republic bru@ujv.cz

4 Kamil Tucek European Commission

kamil.tucek@ec.europa.eu

5 F. Gauche France francois.gauche@cea.fr

6 Forgeron France thierry.forgeron@cea.fr

7 Devictor France nicolas.devictor@cea.fr

8 C. Latge France christian.latge@cea.fr

9 Le Coz France pierre.le-coz@cea.fr

10 Storrer France Francois.STORRER@cea.fr

11 Antonio Sureda Germany antonio.sureda@grs.de

12 Bruno Merk Germany b.merk@hzdr.de

13 Andrej Zeman IAEA A.Zeman@iaea.org

14 Sara Bortot Italy sara.bortot@mail.polimi.it

15 Alessandro Gessi Italy alessandro.gessi@enea.it

16 Edoardo Corsetti Italy corsetti@rse-web.it

17 Tai Asayama Japan asayama.tai@jaea.go.jp

18 Sun-rock Choi Republic of Korea choisr@kaeri.re.kr

19 Frodobertus Klaassen Netherlands klaassen@nrg.eu

20 Ferry Roelofs Netherlands roelofs@nrg.eu

21 Peter Wolniewicz Sweden peter.wolniewicz@physics.uu.se

22 Janne Wallenius Sweden janne@neutron.kth.se

23 Sergii Fomin Ukraine sfomin@kipt.kharkov.ua

No Name of participant Country Email address

24 Bernard VRAY France bvfcsarl@yahoo.fr

25 Stefano Monti IAEA S.Monti@iaea.org

26 Morello Sperandio France morello.sperandio@areva.com

27 A. P. Sorokin Russian Federation sorokin@ippe.ru

28 A. V. Morozov Russian Federation sas@ippe.ru

29 Jean Lou Perrin France jean-lou.perrin@cea.fr

30 Prabhat Kumar BHAVINI pkumar@igcar.gov.in

31 S.C. Chetal IGCAR chetal@igcar.gov.in

32 P. Chellapandi IGCAR pcp@igcar.gov.in

33 D. Sunil Kumar IGCAR sunil@igcar.gov.in

34 T. Satyasheela IGCAR sheela@igcar.gov.in

35 G. Raghu Kumar IGCAR grk@igcar.gov.in

36 P. Puthiyavinayagam IGCAR vinayaga@igcar.gov.in

37 V. Rajan Babu IGCAR vrb@igcar.gov.in

38 R. Vijayashree IGCAR rviji@igcar.gov.in

39 S. Raghupathy IGCAR rags@igcar.gov.in

40 V. Balasubramaniyan IGCAR bala@igcar.gov.in

41 K. Madhusoodhanan IGCAR madhu@igcar.gov.in

42 T. Jayakumar IGCAR tjk@igcar.gov.in

43 B.K. Panigrahi IGCAR bkp@igcar.gov.in

44 V. Karthick IGCAR karthik@igcar.gov.in

45 Rajesh Ganesan IGCAR rajesh@igcar.gov.in

46 K. K. Rajan IGCAR kkrajan@igcar.gov.in

47 P. Selvaraj IGCAR pselva@igcar.gov.in

48 R. Sridharan IGCAR rsn@igcar.gov.in

49 V.A Suresh Kumar IGCAR skumar@igcar.gov.in

50 B. K. Sreedhar IGCAR bksd@igcar.gov.in

58

No Name of participant Country Email address

51 V.Prakash IGCAR prakash@igcar.gov.in

52 K. Velusamy IGCAR kvelu@igcar.gov.in

53 V. Vinod IGCAR vvinod@igcar.gov.in

54 E. Hemanth Rao IGCAR hemanth@igcar.gov.in

55 Sanjay Kumar Das IGCAR skd@igcar.gov.in.

56 B.K. Nashine IGCAR bknash@igcar.gov.in

57 P. Muralidharan IGCAR murali@igcar.gov.in

58 C. P. Reddy IGCAR cpreddy@igcar.gov.in

59 S.B. Degweker BARC degweker@barc.gov.in

60 Shri Anil Bajpai, ThPD BARC

61 D. K. Mohapatra SRI, AERB dina@igcar.gov.in

62 M. Krishnamoorthy BHAVINI mkrishnamoorthy@igcar.gov.in

63 T.K. Mitra BHAVINI tkmitra@igcar.gov.in

64 S. Narasimhan BHAVINI snsimhan@igcar.gov.in

65 P. M. Jagadeesh BHAVINI jagadeesh@igcar.gov.in

66 Obli BHAVINI obuli@igcar.gov.in

67 Y.V.R. Kotteswara Rao BHAVINI yvrkrao@igcar.gov.in

68 A. Ananth BHAVINI aananth@igcar.gov.in

69 G.V.S. Hemantha Rao NFC hemanth@nfc.ernet.in

70 A. V. Satish NPCIL avsathish@npcil.co.in