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P4T-J-222 COMPONENT FAILURE DATA COLLECTION AND ANALYSIS FROM JET AND TLK OPERATING EXPERIENCE

CIATTAGLIA Sergio, T. PINNA(1) G. CAMBI(2) A. LO BUE(1) S.KNIPE(3) J.ORCHARD(3) R. PEARCE(3) U. BESSERER(4)

(1) ENEA FUS/TEC Via E.Fermi 45, 00044, Frascati (Rome), Italy (2) University of Bologna, Physic Department Via Irnerio 46, 40126 BO, Italy (3) JET Joint European Torus Culham, UK (4) FZK, Tritiumlabor, Postfach 3640, D-76021 Karlsruhe

Objective of the activity was to develop a fusion specific component failure database with data coming from operating experiences gained in the Joint European Torus (JET) for the Vacuum and Active Gas Handling Systems (AGHS) and in the Tritium Laboratory of Karlsruhe (TLK). Number of failures/malfunctions occurred in the years of operations, failure modes and, where possible, causes and consequences of the failures, were identified, as well as, whole sets of components, which the anomalies are related to. Components were classified and counted in order to find out amount of components, related operating hours and related demands to operate (for components operating in intermittent way). Main reliability parameters (such as failure rate and corresponding standard errors and confidence intervals) associated to the components were estimated too. 130 failures on a set of 6259 components were pointed out in the AGHS. About 50 different failure rate values were determined (e.g. Bellow pumps-Fail to start 3.1E-5 1/h). 600 failures on a set of 4012 components were pointed out for leaks into JET torus. About half of them were detected after vacuum intervention or/and planned shutdowns, when JET machine was leak tested. The remaining leaks occurred during machine operation. They induced, in about 90 cases, the stop of machine operation. In the remaining cases, the continuation of the operation was possible because either the leaks were small and not interfering with the experimental program or it was possible to fix the leak repairing during operation or, isolating the leak till the next shutdown. About 80 failure rate values were determined for vacuum system (e.g.: Windows-Leak/Rupture 3.4E-6 1/h). 52 failures on a set of 584 components were pointed out for TLK. Even if at a reduced scale respect to JET data, operating data can be considered of some interest for statistical evaluation (e.g.: Catalysts-Heater failure 7.2E-7 1/h). It has to be highlighted that component failure rates here evaluated are: in very good agreement with the corresponding ones existing in literature for similar applications (e.g.: nuclear power plants); one of the most consistent set of data in the field of fusion facilities, both for amount of components treated and for total number of operating hours; very useful in support of safety assessment and for availability/reliability analyses of fusion machines/plants.


Corresponding Author:

CIATTAGLIA Sergio

EFDA CSU Garching, 85748 Garching bei München, Germany

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-232 COLLECTION AND ANALYSIS OF OCCUPATIONAL RADIATION EXPOSURE DATA RELATED TO JET OPERATIONS

Patel Bharat, A. NATALIZIO (1), M. T. PORFIRI (2)

(1)ENSAC Associates Ltd.,87 Markland Dr. Etobicoke, Ont. M9C 1N4, Canada (2)ENEA FUS/TEC - Thermonuclear Fusion Unit, Nuclear Fusion Technologies,Via Enrico Fermi 45, I-00044, Frascati (Rome), Italy

Comprehensive and detailed estimates of occupational radiation exposure (ORE) for ITER are still in progress. Once a detailed design and operating plan emerges, such estimates will be undertaken. In the interim, however, ORE experience from tokamaks, such as JET, represents a valuable source of information to assist in the development of future estimates, and, possibly, in validating them. Accordingly, the JET ORE experience was reviewed, for the first time, from an ITER perspective. A systematic review of JET annual ORE data from 1988 to 2002 was performed to identify trends and relationships that would be useful for future ITER ORE estimates. The preliminary results highlight a number of areas of interest to ITER and, the findings justify a continued detailed study of the ORE experience of JET and other tokamaks. The analysis of the JET experience to date has yielded the following results and conclusions: (1) on average, the machine was in the shutdown state almost 50% of the time; (2) while most of the shutdown time was planned for maintenance or modifications, about 18% of the total shutdown time, on average, was due to unplanned machine interventions; (3) about two thirds of the JET annual dose, on average, was accrued by the maintenance staff; (4) about 15% of the average, annual worker dose was accrued during the machine operating state; (5) annual worker doses were significantly reduced after the implementation of the ALARP policy in 1998. This demonstrates the importance of remote-handling equipment in reducing worker doses; (6) the long-term averages of the collective and individual worker doses are 96 p-mSv/a and 0.54 mSv/a. This compares with the ITER design targets (500 p-mSv/a and 2 mSv/a, respectively); (7) moreover, the post-1997 averages (ie after DTE1) are lower, ie, average collective annual dose was 30 p-mSv/a and the average individual dose was 0.06 mSv/a; (8) on average, since 1997, the tritium dose has been in the order of one percent of the total worker dose. This result shows that use of remote handling measures following the significant use of tritium as part of a more stringent ALARP policy, and the practical radiation protection measures used at JET (eg use of ventilation controls and airline pressurized suits), worked well at reducing the tritium dose to a small percentage of the total dose.


Corresponding Author:

Patel Bharat

Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3EA, UK

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-244 TFTR OCCUPATIONAL RADIATION EXPOSURE DATA COLLECTION AND ANALYSIS

PINNA Tonio, A. NATALIZIO (1) J. D. LEVINE (2)

(1) ENSAC Associates Ltd., 87 Markland Dr. Etobicoke, Ont. M9C 1N4 (2) DOE Princeton Plasma Physics Laboratory P.O. Box 451 Princeton, New Jersey 08543

Occupational radiation exposure (ORE) experience from existing fusion facilities represents a valuable source of information to assist in the development of ORE estimates for future machines, such as ITER, and, possibly, in validating them. Accordingly, the worker dose experience from TFTR was of particular interest, not only because of the machine size, but also because of its D/T operation and because it saw the complete cycle of the plant life, decommissioning phase included. The TFTR ORE experience was reviewed from an ITER perspective. A systematic review of TFTR worker dose data was performed to identify trends and relationships that would be useful for future ITER ORE estimates. Although more could have been done with additional data, particularly missing information, the analysis has yielded some useful results and findings. The first finding is that once the machine becomes activated, collective worker dose will be proportional to machine shutdown time, even if the proportionality constant has a high degree of variability. For TFTR, the ratio of exposed workers to monitored workers was almost one half and, the shutdown time, when most of the dose is accrued, was, on average, about 60% of the year. The second finding is that the application of an ALARA policy, as was applied in the case of TFTR in 1993, can have a significant impact on the worker collective dose. The third finding is that although the major portion of TFTR worker dose was accrued by maintenance workers, during shutdown periods, a significant portion (up to 20%) was accrued by non-maintenance work groups during operating periods (excluding plasma operation). Therefore, operating doses cannot be neglected. Finally, after several years of D/T operation, it can be concluded that TFTR tritium doses did not contribute significantly to the total worker dose.


Corresponding Author:

PINNA Tonio

ENEA CRE Frascati, Via E. Fermi 45, 00044, Frascati (Rome), Italy

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-260 FACTORS AFFECTING THE INHALATION DOSE FROM TRITIATED DUST AND FLAKES

DI PACE Luigi, PATEL Bharat

JET Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3EA, UNITED KINGDOM

Tritiated dust and flakes have been produced in JET during D-T campaigns. Measurements carried out so far showed some atypical radiological effects, in particular a very large T concentration, and a larger dose for particulates intake than for a similar HTO intake. The large tritium content in the dust was not expected, but the higher dose per unit intake was foreseen and is a consequence of the particulate nature of the material and of the expected higher biological retention. A study was carried out on the radiation hazards of tritiated dust and flakes from JET and other tokamaks. The particle size plays an important role in the deposition pattern in the human respiratory tract. The present study pointed out the most significant size dependent parameters influencing particle mobility and deposition in the human respiratory tract. One of those is the Activity Median Aerodynamic Diameter (AMAD). Measurements on JET 1999 samples showed that the AMAD was ~4 ìm, in the range of respirable aerosols. Larger particles are mostly deposited and removed in less than one day in the anterior nasal passage and remaining airways of the head and the neck. Smaller particles are retained and cleared in time periods ranging from few hours, in the bronchial and bronchiolar regions, up to years, in the alveoli. In the latter, the only clearing mechanism of tritium is its absorption into the lung serum. This phenomenon is a two-stage process (dissolution + uptake) and it is an important factor to be assessed in the calculation of dose and dose conversion factors. The dissolution can be reproduced by in-vitro tests. Some in-vitro tritium dissolution tests were carried out on carbon dust collected in JET, but the results were difficult to compare with other data available in literature. From the literature related to other in-vitro dissolution experiments, it has been possible to formulate some recommendations for the next in vitro experiments to be carried out on JET dust. -Better dust characterisation (chemical composition, size, surface specific area, T distribution inside the dust); -Longer tests (at least 3 months) on two classes of dust dimensions (CMD=1 µm and 5 µm); -Evaluation of tritium dissolution behaviour from particles since the very beginning of tests; -Perform beta radiation measurements on dust to be compared to the one by calculation assuming tritium homogeneous distribution; -Determine the chemical composition of dust in addition to C and H.


Corresponding Author:

DI PACE Luigi

ENEA C.R. Frascati, via Enrico Fermi 45, -00044, Frascati (Rome), ITALY

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-271 THE EUROPEAN POWER PLANT CONCEPTUAL STUDY

Maisonnier David, Cook I. (2), Boccaccini L. (3), Bogusch E. (4), Broden K. (5), Di Pace L. (6), Forrest R. (2), Giancarli L. (7), Hermsmeyer S. (3), Nardi C. (6), Norajitra P. (3), Pizzuto A. (6), Sardain P. (1), Taylor N. (2), Ward D. (2)

1. EFDA CSU Garching 2. Association Euratom-UKAEA 3. Association Euratom-FZK 4. EFET-Framatome ANP GmbH 5. Studsvik Radwaste AB 6. Association Euratom-ENEA 7. Association Euratom-CEA

Selected also for oral presentation O4B-J-271

Within the European Power Plant Conceptual Study 4 fusion power plant “models” have been developed. They are all based on the tokamak concept and they have the same net electrical power output, 1500 MWe. In many other respects, corresponding to different degrees of extrapolation both in physics and in technology, the 4 models differ. For the models A and B, based on “limited” extrapolations, significant efforts have been devoted to ensure the consistency of the concepts developed, which has required better clarification of the physics basis, development of adequate divertor concepts and the proposal of a novel blanket segmentation. PPCS model A is based on the WCLL blanket and on a water-cooled divertor concept. PPCS model B is based on the HCPB blanket and on a helium-cooled divertor concept. A low-activation martensitic steel has been considered as the main structural material for both concepts. Two key innovative developments are worthy of especial note and they will be described in some detail: (i) the development of a scheme for the scheduled replacement of the internal components, which results in a plant availability in excess of 75%; (ii) a conceptual design for a helium-cooled divertor able to resist heat-loads of 10 MW/m2. For each of the more advanced concepts, an advanced physics scenario has been identified and combined with advanced blanket concepts that allow higher coolant temperatures and, consequently, higher thermodynamic efficiencies of the power conversion systems. Together, these allow plants of smaller dimensions to be considered. For all models, systems analyses were used to integrate the plasma physics and technology constraints, together with other considerations such as unit size and availability, to produce self-consistent plant parameter sets with approximately optimal economic characteristics. In the PPCS Models, the favourable, inherent, features of fusion have been exploited to provide substantial safety and environmental advantages. The broad features of the conclusions of previous studies have been confirmed for the new models and demonstrated with increased confidence. Finally, the PPCS study has highlighted the need for specific design and R&D activities in addition to those already underway within the European long term R&D programme. These are needed to develop advanced physics scenarios and to confirm the feasibility of the proposed maintenance scheme and of the proposed divertor concepts.


Corresponding Author:

Maisonnier David

EFDA-CSU Garching, Boltzmannstr. 2, D-85748 Garching

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-273 INTRA ANALYSIS OF WET BYPASS TRANSIENTS INCLUDING TRITIUM

zeMichael Yitbarek, Ove Edlund

Studsvik Nuclear AB SE-611 82 Nyköping

An INTRA model of the recent ITER facility design has been developed, based on data from the Safety Analysis Data List (SADL-4.0.0), to simulate postulated accident events specified in the Accident Analysis Specifications (AAS-4.beta.1). This paper summarizes the model and simulation results of one large wet bypass event. The large wet bypass event simulates a multiple First Wall (FW) pipe break initiated by failure of a horizontal penetration line, which connects the Vacuum Vessel (VV) to the Port Cell (PC). It is postulated that all the 54 FW modules, around the inboard and outboard toroidal circumference of the machine, are damaged discharging coolant water at a high flow rate directly into the VV. Loss of offsite power is assumed to coincide with the initiating event, which stops all pumps and force the coolant loops to continue in natural circulation mode. The main objectives of this study are: - to analyse the pressure and temperature transients in the VV, VV pressure suppression system (VVPSS), drain tank, PC and gallery, - to show that mobilized tritium is adequately confined during this kind of accident, - to validate performance requirements of the suppression tank venting system (ST-VS), the port cell, the vacuum breakers and the standby vent detritiation system (S-VDS). The simulation result shows that highest pressure, 167 kPa, is attained in the VV, which is below the design pressure of 200 kPa. The tritium release to the environment is estimated to 3.31 g, which is also much lower than the highest permissible release, 90 g. The vacuum breakers have maintained the pressure in the gallery, cryostat space room and port cell at equilibrium, fulfilling its performance requirement and the S-VDS has kept the pressure in these rooms sub-atmospheric. Furthermore, the result shows that the VVPSS pressure is kept sub-atmospheric by actuating the ST-VS, thus preventing tritium and other radioactive particles being released to adjacent rooms. INTRA is one of the reference codes, specified by the ITER Joint Central Team and the European Community, for safety analyses of Tokamak type fusion reactors. In this study, INTRA-mod6.beta version is used. This special version is developed to enhance the previous INTRA version by including tritium transport simulation.


Corresponding Author:

zeMichael Yitbarek

Studsvik Nuclear AB, SE-611 82 Nykoping

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-317 ACCESSIBILITY EVALUATION OF THE IFMIF LIQUID LITHIUM LOOP CONSIDERING ACTIVATED EROSION/CORROSION MATERIALS DEPOSITION

Nakamura Hiroo, Takemura Morio(1), Yamauchi Michinori (1), Fischer Ulrich (2), Ida Mizuho (1), Mori Seiji (3), Nishitani Takeo (1), Simakov Stanislav (2), Sugimoto Masayoshi(1)

(1) Japan Atomic Energy Research Institute, 2-4 Shirakata-Shirane, Tokai-mura, Ibaraki-ken, 319-1195 Japan (2) Association FZK-Euratom, FZK,76021 Karlsruhe, Germany (3) Kawasaki Heavy Industries, Ltd., Tokyo 136-8588, Japan

International Fusion Materials Irradiation Facility (IFMIF) is a deuteron-lithium (Li) stripping reaction neutron source for fusion materials testing. A liquid Li target has been designed to produce intense high energy neutrons for the material irradiation up to 50 dpa/y by 10 MW of deuterium beam deposition. Since the activated Li target materials by neutron irradiation are distributed due to erosion/corrosion process under liquid Li flow, accessibility during hands-on maintenance around the Li loop pipings will depend on activation level of the deposition materials although the target assembly is designed as fully remote maintenance component. This paper presents an evaluation of accessibility of the Li loop pipings considering activated corrosion product. Activation level is calculated by the ACT-4 of the THIDA-2 code system. High energy cross section above 15 MeV in the ACT-4 is modified using IEAF-2001 data. In this calculation, target material is stainless steel 316. Area of the erosion/corrosion in back wall is 100 cm2 same as deuterium beam foot print on a Li free surface. The erosion/corrosion rate is selected as 1x10-6 m/y using data from FMIT project. Decay of radioactivities of the main radioisotopes in the IFMIF Li loop has been calculated after shutdown up to 1000 days. Main radioactive species above 1 Ci/cc after 1 week from the shutdown are Co-58, Mn-54, Co-56, Co-57,Fe-59 and Cr-51. Total radioactivity is about a few tens Ci/cc. Dose rate around the Li loop after one year IFMIF operation has been calculated assuming uniform deposition on Li loop surface area of 33 m2 and deposition rate of 1,10,100%. Permissible level for hands-on maintenance is selected as 1x10-5 Sv/hr. As the results, in case of 10% to 100% deposition, close maintenance and hands-on maintenance are difficult. In case of 1 % deposition after 1 week from shutdown, close maintenance work 8 cm to the Li loop is possible. Also, after 1 month, hands-on maintenance becomes possible. Design issues on removal of the activated radioactive materials and components maintenance will be also discussed.


Corresponding Author:

Nakamura Hiroo

Office of Fusion Materials Research Promotion, Tokai Research Establishment, Japan Atomic Energy Research Institute, 2-4 Shirakata-Shirane, Tokai-mura, Ibaraki-ken, 319-1195 Japan

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-336 AVAILABILITY OF LITHIUM IN THE CONTEXT OF FUTURE D-T FUSION REACTORS

FASEL Damien, MQ.TRAN

Centre de Recherches en Physique des Plasmas (CRPP),EPFL/SB,PPH 280,CH-1015 LAUSANNE

Since several years, one of the main worldwide preoccupations is the future energy supply for all countries. Thus, one of the challenges of this century will be to define and to choose a coherent strategy to supply energy for all countries, taking into account the following aspects: the total energy demand will continue to increase, the environmental damage (pollution, greenhouse gases, CO2 etc.) will require clean energy sources, and finally, the concept of sustainable development will have to be implemented. In the light of those aspects, the energy which could be generated by thermonuclear reactions, will be of particular interest and will certainly play an important role. A future fusion reactor will use Deuterium and Tritium as fuel. The former is found in abundance in seawater, while the latter is not available naturally due to its short lifetime. Present developments in the context of fusion propose generating Tritium in a blanket based on the reaction n + Li6 -> T+He. This paper will review the issue of the reserves and resources of Lithium as the fuel for fusion energy within the context of sustainable development. We describe the available reserves and resources and their chemical forms. We analyse the present production and consumption for industrial uses, completing this overview by referring to a life cycle analysis in the case of the lithium production from brine. The particular case of lithium extraction from seawater is also part of this study. A possible method of extraction based on solar ponds and filtering resins is presented and discussed, as well as a new promising investigation using ion-sieve manganese oxide. Taking into account the previous results, both from the present consumption and its forecasted evolution, as well as the additional requirements estimated for fusion plants, we analyse several scenarios for future lithium consumption. The impact of each of them on the lithium reserves will be evaluated. As a conclusion we discuss, on the basis of these different hypothesis considered, the lithium availability over the next centuries.


Corresponding Author:

FASEL Damien

Centre de Recherches en Physique des Plasmas (CRPP), EPFL/SB,PPH 280, CH-1015 LAUSANNE

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-342 EFFECT OF ACTIVATION CROSS-SECTION UNCERTAINTIES IN SELECTING STEELS FOR THE HYLIFE-II CHAMBER TO SUCCESSFUL WASTE MANAGEMENT

Javier Sanz (1, 2), Oscar Cabellos (2) Arturo Rodríguez (1) Susana Reyes (3) Jeffery F. Latkowski (3)

(1)UNED. ETS Ingenieros Industriales. Dept. Ingeniería Energetica. C/Juan del Rosal, 12. 28040 Madrid, Spain. (2)Instituto de Fusión Nuclear (UPM). Madrid, Spain. (3) Lawrence Livermore National Laboratory. California, USA.

In an inertial fusion energy (IFE) thick-liquid chamber design such as HYLIFE-II, a molten-salt is injected between the fusion explosions and the chamber walls to attenuate neutrons and protect the structures from radiation damage. The recent ARIES-IFE study has provided an updated assessment of the choice of chamber structural material for HYLIFE-II. In addition to the initial choice of SS304, other alternate steels, austenitic and ferritic, have been proposed. Activation and waste management analyses have been an important part of the evaluation, focusing on the shallow land burial (SLB) option. It was recognized that further analysis is needed to determine allowable impurity levels and optimized compositions for low-activation fusion applications. This paper assesses the long-lived activation properties of some unintended and intended alloying constituents considered critical for producing real low-activation steels in the neutron environment of the HYLIFE chamber. Thus, the concentration limits (CL) for these elements according to SLB restrictions will be estimated. These CL quantities are considered as random variables whose uncertainties due to activation cross sections uncertainties will be evaluated. Concentration limits for recycling options will be also estimated. A comprehensive methodology to compute uncertainties on activation calculations have been developed and implemented in the activation code ACAB. A sensitivity-uncertainty analysis approach and a Monte-Carlo based methodology have been used. The effect of each reaction cross-section separately, and the synergistic/global effect of the complete set of cross-sections uncertainties is assessed. The latest available activation cross-section data are employed, and useful uncertainty information (variances and co variances) is collected, processed and used. The results will allow answering if HYLIFE-type reactors with steels of present-day technology could be acceptable from the waste management perspective or otherwise more advanced medium or near-term options could be attractive alternatives. Also, the significance of the MFE steels-materials program to IFE-HYLIFE applications could be better evaluated, on account of the dependence of the results on the neutron energy spectrum. Furthermore, the quality of the current nuclear data for the specific IFE application here considered is analyzed, and if necessary new cross sections measurements or evaluations will be proposed.


Corresponding Author:

Javier Sanz (1, 2)

Universidad Nacional de Educación a Distancia. ETS Ingenieros Industriales. Dept. Ingeniería Energética. C/Juan del Rosal, 12. 28040 Madrid, Spain

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-357 EVALUATION OF FUSION STUDY FROM SOCIO-ECONOMIC ASPECTS

Konishi, Satoshi, Kunihiko Okano(1) Yuichi Ogawa(2) Shunichiro Nagumo(3) Koji Tokimatsu(4) Kenji Tobita(5)

(1)Central Research Institute of Electric Power Industry (2)The University of Tokyo (3)The Japan Research Institute (4)Research Institute of Innovative Technology for the Earth (5)Japan Atomic Energy Research Institute

Effects of the research and development of fusion energy, “Value for Money” is requested to be evaluated and explained by funding government and public for justification of its investment, in comparison with other energy researches, scientific research projects or, in some cases, with other technology projects or investment, for cost effectiveness. The authors have attempted to evaluate a fusion development project in Japan from various aspects of Socio-economic study of fusion for its impacts on economy, public and society. Major part of such impact occurs outside of the market mechanism, and methodology developed for the evaluation of “Externality” was used. In the consideration of the impact of energy development, there are many other influence paths were taken into account. Here four categories were identified as such mechanisms. First two effects are direct economic effect by the research investment spent for purchase and employment, and the second is its extension observed as the growth of local community and its economy. These effects are always seen, and evaluation procedure is established. They are however seen in other construction projects and are not specific for scientific research. The third effect is an improvement of technical capability of the industry stimulated by the development of novel scientific equipments. Technical spin-offs are regarded as part of it as Externality, however technical investments for research that does not yield immediate commercial products would benefit industrial society in various manner, and the results usually appears slow and occasional. Such an impact can be evaluated from the markets of technical products, but in the present case study, contribution of the research activity and investment attributed to such markets are usually several % of the total sales, that is almost same level of funding that industries spends for developments. The last, and largest impact that fusion research can result is through energy supply, but it only appears several decades future, and should be deducted by discount rate and success probability. Value of the fusion research at present age cannot be estimated as actual benefit, but as the effectiveness in the policy scenario of energy supply under environmental constraints in global scale. The presentation will show an example of methodology to evaluate fusion research, including comparison with other energy development and scientific projects.


Corresponding Author:

Konishi, Satoshi

Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto, 611-0011 Japan

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-373 PROGRESS IN THE DEVELOPMENT OF A PIE-PIT FOR THE ITER TOKAMAK

Jesus Izquierdo (1), Neill P. Taylor (2) Javier Dies (1) Jeronimo Garcia (1) Ferran Albajar (1)

(1) FEEL - Fusion Energy Engineering Laboratory Universitat Politècnica de Catalunya Av. Diagonal 647, 08028 Barcelona, Spain (2) EURATOM/UKAEA Fusion Association Culham Science Centre Abingdon, Oxfordshire, OX14 3DB, United Kingdom

The ITER Postulated Initiating Event and Potential Impacts Table (ITER PIE-PIT) is a matrix that shows the potential impacts on ITER Plant States from postulated initiating events. The PIE-PIT for the ITER Tokamak was developed leading to the identification of 162 event sequences with impact on the Plant State [1]. In this paper, a ranking scheme has been applied to the tokamak matrix in order to sort the 162 event sequences by risk, frequency and severity. The safety-related parameters or factors influencing the ranking have been the frequency of every PIE and its severity, the severity of the final Plant State and the reliability of confinement barriers and other safety related systems. Preliminary results denote a strong presence at the top of the ranking of the sequences involving ingress of coolant (water and cryogenic coolant) as well as sequences involving failure of fuelling system components. Comparison with those sequences detailed in the ITER Generic Site Safety Report (GSSR) shows adequate coherence. The implementation of the PIE PIT methodology for ITER allows to rank priorities of accident sequences to be assessed in next steps, from the probabilistic and deterministic points of view. For this reason, sequence simulation software is being developed at Fusion Energy Engineering Laboratory with a strong focus on plasma-wall interactions [2] and further in-vessel phenomena. The main feature of this software will be its modular oriented programming in order to analyze several aggravating failures and accident sequences. [1] 'Development of a PIE-PIT for ITER and selection of bounding event sequences', N.P. Taylor, UKAEA FUS 489, rev. 2, May 2003 [2] 'Development of time dependent safety analysis code for plasma anomaly events in fusion reactors', T. Honda, H-W. Bartels, N.A. Uckan, Y. Seki, T. Okazaki, Journal of Nuclear Science and Technology, vol. 34, n. 3, p. 229-239, March 1997


Corresponding Author:

Jesus Izquierdo (1)

FEEL - Fusion Energy Engineering Laboratory, Escola Tèc. Sup. d'Enginyeria Industrial de Barcelona, Dep. de Física i Enginyeria Nuclear, Universitat Politècnica de Catalunya, Av. Diagonal 647, Pav. C, 08028 Barcelona, Spain

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-380 GLOBAL ENERGY MODEL WITH FUSION

Lechon Yolanda, Cabal H. Varela M. Eherer C. Baumann M. Düweke J. Hamacher T. Tosato G.

CIEMAT Avda Complutense 22 28040 Madrid (Spain) TUG Technikerstraße 4 8010 Graz (Austria) IPP Boltzmannstr. 2, D-85748 Garching bei Muenchen (Germany) EFDA Close Support Unit Boltzmannstr. 2, D-85748 Garching bei Muenchen (Germany)

The global energy system is expected to make a complete shift in the 21st century, away from fossil fuels to either renewable or new nuclear technologies. The rational behind the shift are numerous: resource depletion, environmental concerns, especially global warming, and unacceptable geo-political frictions. Fusion might become a corner stone of the future energy system. The construction and successful operation of ITER and the successful qualification of materials for future fusion plants is a necessary condition to reach this goal. A consortium of European energy and fusion research laboratories has been formed to design and operate a single region global energy model based on the TIMES model generator, to elaborate the possible role of fusion in the future energy system. The TIMES model generator has been developed by expert contractors under the IEA Energy Technology Systems Analysis Programme (ETSAP). Using TIMES, a single region global model has been constructed including fusion as an energy option. Background of the model is a detailed bottom-up description of the complete energy system starting from mining process up to the various demand sectors. The model dynamics is determined by an optimisation process, in which all costs are minimised. Future costs are discounted according to a certain discount rate. Within the Socio Economic Research on Fusion (SERF) programme guided by EFDA a consortium between CIEMAT, TU Graz, ENEA and IPP was formed. The paper will present the first attempts to set-up a single region global model and will present first results of the model. The results will be discussed in comparison to former results of the SERF activity, which were limited in geographic scope to Europe and India. In future the model will be expanded to a multi-region global model which is now being developed by expert team under an EFDA contract.


Corresponding Author:

Lechon Yolanda

CIEMAT Avda Complutense 22 28040 Madrid Spain

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-400 DUST EXPLOSION HAZARD IN ITER: EXPLOSION INDICES OF FINE GRAPHITE AND TUNGSTEN DUSTS AND THEIR MIXTURES

Denkevits Andrey, Dorofeev Sergei

Forschungszentrum Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

The work addresses one of the ITER safety problems: the explosion hazard of the dusts produced inside the ITER vacuum vessel as a result of plasma-wall interaction. The administrative guidelines assume for ITER several hundred kilograms of carbon, tungsten, and beryllium dusts to be accumulated there. In case of air and/or steam ingress into VV the dusts can be mobilized to form a dust-air cloud capable to explode generating the pressure loads. Standard method of 20-l-sphere is used to measure the explosion indices of fine graphite and tungsten dusts and their mixtures as ITER-relevant dusts. Maximum explosion overpressures and rates of pressure rise are measured in wide concentration ranges. Lower explosion concentration limits (LEL) are measured as a function of the graphite dust characteristic particle size, of 1-micron tungsten dusts, and graphite/tungsten dust mixtures. Explosibility of the dusts is evaluated. Three graphite dusts are tested differing in their fineness: 4-micron, 25 micron, and 40 micron characteristic particle size. The dust particle size is shown to have a profound effect on the explosion characteristics: the finest tested dust features the highest maximum overpressure (6.6 bar) and rate of pressure rise (250 bar/s), and the lowest LEL (70g/m3). It can be exploded by 2 kJ energy igniters, i.e. it can be classified as explosive. Three tungsten dusts tested have 1-micron, 5-micron, and 12-micron particle size. The two coarser dusts do not appear to explode in 250-7000 g/m3 concentration range. The finer dust explodes from 450 to 7500 g/m3 (the highest tested concentration). The maximum overpressure of 4.7 bar and maximum rate of pressure rise of 260 bar/s are measured at 5000 g/m3 concentration. The mixtures are composed of 4-micron graphite and 1-micron tungsten dusts. The four mixtures are tested differing in the composition: having tungsten-to-graphite molar ratio W/C=1/30, 1/4, 1/1, and 3/1. To study conservative cases, the mixtures are tested at the optimum concentrations at which the dust combustion consumes all the oxygen in air producing maximum overpressures and rates of pressure rise. The maximum overpressure decreases slightly with increasing tungsten content. The maximum rate of pressure rise has a pronounced peak of 360 bar/s at W/C=1/1, i.e. this mixture burns faster than both pure graphite and pure tungsten dusts (250 bar/s) alone. Nevertheless, even the most explosive mixture belongs to the mildest explosion class St1.


Corresponding Author:

Denkevits Andrey

Forschungszentrum Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-401 ECONOMIC ANALYSIS OF FDS FUSION POWER REACTORS

Desuo Huang, Yican Wu, Liqin Hu, Delin Chu and FDS Team

P.O.Box 1126,Hefei,Anhui,230031,People’s Republic of China

The series of fusion-driven systems FDS-I/II are studied by The FDS?Fusion Design Study?Team at Institute of Plasma Physics, Chinese Academy of Sciences. FDS-I is a fusion-driven sub-critical reactor with a fission blanket, which could perform multi-functions including breeding fissile nuclear fuel, transmuting long-lived wastes, producing tritium for fusion fuel cycling and generating power etc.,on the basis of lower core plasma physics and engineering parameters. FDS-II is a pure fusion reactor, which could demonstrate commercially efficient generation of electricity. It is very important to make the systems be economically attractive. The cost of electricity (COE) is one of key parameters to be used to evaluate the economy of power reactors. However, the previous models for economic analysis didn’t include the benefit of transmutation of wastes and breeding of fissile nuclear fuel and do not apply to FDS-I power reactor. In this contribution, a model describing the economics characteristics of FDS-I power plant has been developed, where the benefit of transmuted wastes and produced fuel is included. COE is considered as a function of the weight of bred fissile fuel, transmuted wastes and blanket energy-gain factor as well as the weight and size of the reactor system. The economy of FDS-I is analyzed with the system analysis code developed on the basis of the developed model. For the purpose of comparison, the economy of FDS-II and other fusion power reactors such as ARIES-I-IV/AT etc. are also calculated and analyzed based on the available designs. Some recommendations for improvement of the designs of FDS power reactors are made based on the economic optimization analysis.


Corresponding Author:

Desuo Huang

P.O.Box 1126,Hefei,Anhui,230031,People’s Republic of China

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-414 RELIABILITY ANALYSIS OF BLANKET MODULES OF FDS

P.Liu, Y. C. Wu(1), Q. Y. Huang(1) and FDS Team(1)

(1)Institute of Plasma Physics, Chinese Academy of Sciences,P. O. Box 1126, Hefei, Anhui 230031,CHina

The series of fusion-driven systems FDS-I/II are studied by the FDSiFusion Design StudyjTeam at Institute of Plasma Physics, Chinese Academy of Sciences. FDS-I is a fusion-driven sub-critical reactor with a fission blanket, which could perform multi-functions including breeding fissile nuclear fuel, transmuting long-lived wastes, producing tritium for fusion fuel cycling and generating power etc, on the basis of lower core plasma physics and engineering parameters. FDS-II is a pure fusion reactor which could demonstrate commercially efficient generation of electricity. It is very important to assess the reliability and availability of the blanket modules as the key components of FDS-I/II. However it is difficult to draw convincing conclusions without suitable reliability data from operation experiences since real fusion power systems are not available in the world. A new approach is adopted by combining the probabilistic fracture mechanics (PFM) method and probabilistic safety assessment (PSA) method in this contribution. The PFM method, which is regarded as an appropriate method in evaluating the reliability of structural components, is used to evaluate the reliability of walls and pipes in blanket modules by considering material, shape, environmental loads such as radiation, temperature, pressure and so on. The PSA method, which is regarded as an appropriate method in evaluating the reliability of a system, is used to evaluate blanket modules reliability by considering their structure, principle and the relation between walls and pipes. A reliability model and an analysis code for the FDS blanket modules is developed. The reliability of FDS-I/II blanket modules is calculated and analyzed with the developed code. Key aspects of affecting the reliability of FDS blanket modules are specified.


Corresponding Author:

P.Liu

Institute of Plasma Physics, Chinese Academy of Sciences,P. O. Box 1126, Hefei, Anhui 230031,CHina

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-421 FUSION SAFETY STUDIES IN RUSSIA IN 2003.

Kolbasov Boris, M.I. Guseva (1), B.I. Khripunov (1), Yu.V. Martynenko (1), A.M. Zimin (2), S.A. Bartenev (3)

(1) Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia (2) Bauman Moscow State Technical University, 2 Bauman st. 5, 105005, Moscow, Russia (3) Khlopin Radium Institute, 2 Murinskij av., 28, 194021 St. Petersburg

The paper is a review of Russian fusion safety studies performed in 2003. The studies were focused on the behavior of plasma facing materials (C, Be, W and their mixtures) under normal ITER operation and at plasma disruptions, especially on H-isotope accumulation in these materials. Unexpectedly W-sputtering in D-plasma at subthreshold ion energies (5-10 eV) typical for gas divertor was observed at target temperatures of 1170-1600 K. Sputtering yield was (1-5) x 10*-4 atom/ion for different W grades. Below 1250 K the W-sputtering does not take place. The mechanism of the subthreshold W-sputtering (Eth=160-178 eV) based on the potential sputtering of adatoms by slow deuterons has been proposed. At dose increase from 1.5x 10*26 to 4.3x10*26 m*-2, W-sputtering yield reduced up to one-twelfth. Evaporation and droplet erosion are typical for W and Be surfaces at plasma disruption simulation. Large droplets fly away almost perpendicular to the surface. Small droplets fly off in parallel tothe surface or return to it. The D/Be atomic ratio in Be-layers, redeposited under simulation of normal ITER operation, decreases from 0.15 at 375 K to 0.05 at 575 K. Admixed C and W slightly enhance D-retention. H-concentration in Be+C films formed on Be under exposure to C2H3 ion flux (Be/C = ~1/2) at 670 K decreases from 20-25 at.% at irradiation dose of 10*23 m*-2 to 6 at.% at 10*24 m*-2. When Be and C targets were simultaneously irradiated by D-plasma at 670 K, the D/(Be+C) atomic ratio in a Be+C layer dominated by Be (Be/C = ~2.3) at a dose of 10*24 m*-2 was 0.03. Under exposure to a D-plasma, films grow and flake off continuously. H-isotopes do not permeate beyond the mixed layers. Thus, the formation of mixed Be+C layers can protect plasma-facing components from erosion and H accumulation. The D/(Be+C) ratio in mixed Be+C layers on Be and CFC-targets under plasma disruption simulation was 0.019. The radiochemical reprocessing technology for a V-Cr-Ti alloy activated in a fusion reactor was developed and tested. It is based on extraction of V, Cr and Ti freed of activation products from an alloy dissolved in nitric acid. The solution of di-2-ethyl-hexyl-phosphoric acid in a dodecane serves as an extractant. It takes 48 extraction steps to recover V, Cr and Ti down to an effective dose rate <12.5 mSv/h, permitting the refabrication of these metals without biological shielding. Cost analysis has shown that the alloy refabrication is preferable against its burial.


Corresponding Author:

Kolbasov Boris

Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-423 CORE CONCEPTUAL DESIGN OF FDS FUSION POWER REACTORS

D.L. Chu, Y.C. Wu, B.Wu, D.S.Huang and FDS Team

Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126, Hefei, Anhui 230031,China

The series of fusion-driven systems FDS-I/II are studied by The FDS(Fusion Design Study)Team at Institute of Plasma Physics, Chinese Academy of Sciences. At the primary stage, FDS-I is designed to seek for early application of fusion energy with lower core plasma requirements. Preliminary core conceptual design for advanced FDS-II, which aims at pure fusion energy application, is presented in this paper. Compared with earlier studies, FDS-II is designed to assess the status-to-be of fusion energy application, and provide operating experience for large-scale commercial electricity production in the future. Based on the reasonable extrapolation of FDS-I designs, the core parameters of FDS-II are designed with : R0=6.0m, a=1.5m, kappa=2.0, delta=0.5 ,BT=7 to 10T, IP=10MA for about 2GW of fusion power. In addition, some physics and engineering parameters such as the fusion triple product n¦ÓT>5¡Á1021 m-3. keV. s, bootstrap current fraction fBS>80%, fusion gain Q>=30, Power density Pden¡Ö5MWm-3, neutron wall loading Pnwall¡Ö3MWm-2, Availability factor favail>0.6 are desired as the key characteristics for FDS-II. The general reactor outline of FDS-II is also determined. As attractive feature of steady-state fusion power plant, advanced reversed shear operation mode is one of important pursuing goals. At the same time, core plasma configuration evolution is simulated by using two-dimensional time dependent free boundary simulation code. Comparison between FDS-I and FDS-II has also been drawn. The results presented here suggest that the extrapolation from FDS-I is reasonable and the core plasma design of FDS-II is self-consistent. Advanced operation mode for FDS-II can be accessed by efforts. FDS-II has potential to be an attractive fusion power plant.


Corresponding Author:

D.L. Chu

Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126, Hefei, Anhui 230031,China

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-452 NEUTRON ACTIVATION AND DOSE RATES MINIMIZATION ON LASER MÉGAJOULE (LMJ) FACILITY

JOYER Philippe, H.P Jacquet, M. Dupont, M. Messaoudi, L. Lachèvre

CEA/Centre d'étude Ile de France, BP 12 91680 Bruyères le Châtel, France Assystem Services, Immeuble central gare, 1 place Charles de Gaulle, 78180 Montigny le Bretonneux

LMJ is the major French project for Inertial Confinement Fusion (ICF) research in France and building construction phase have started in the middle of the year 2003. 240 laser beams will generate energy of 1.8 MJ on a deuterium-tritium target. High 14 MeV neutrons amounts (up to 1 1019 neutrons) will be produced with a few hundred micrograms of deuterium tritium mixture. Neutron generation and tritium use in the targets will induce: - high neutron radiation during a shot, - gamma radiation between shots due to activation of materials - contamination of all in vessel equipments Neutron activation and gamma doses rates induced are a real challenge for facility operation and maintenance. Calculations have been undertaken several years ago using Tripoli and Fispact computer codes. Preliminary results, presented at the 22nd SOFT, have shown rather high dose levels in the experimental area one week after yield shots campaigns and the need to reduce activation in this area. The complexity of the facility requires taking into account the maximum of equipments which will be implemented in the target bay. At present time, most component designs have been introduced in the model and the full capability of the code has nearly been reached. Complementary calculations have been realized on the basis of parametric studies (addition of optimized amounts of boron, composition of alloy including measured impurity levels, addition of shielding…) to determine the best compromise to limit activation. Results will be given and reduction factors of more than one decade, depending on the working area, have been got.


Corresponding Author:

JOYER Philippe

CEA/CESTA BP2 33114 Le Barp, France

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-524 DESIGN EARTHQUAKES FOR ITER AT CADARACHE

Girard Jean-Philippe, G. Grünthal (2), M. Nicolas (1), and EISS Team

(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) GeoForschungsZentrum Potsdam, Germany

Input requirements and assumptions for ITER consider that an infrequent, severe earthquake (called SL-2) which, although unlikely to occur during the lifetime of the facility, is assessed to demonstrate adequate protection of the public. This earthquake is assumed to have a return period of 10,000 years. An investment protection level or inspection level (where all Structures, Systems and Components are safe) with a peak ground acceleration (PGA) at 0.5 m/s2 is also considered. As a basis, orders of magnitude of consequences, if no countermeasures were taken, are given. The European site proposed for ITER is situated in the South of France, 40 km North of Marseille, in a low to moderate seismic area according to the Global Seismic Hazard Map (GSHAP Group 1999). The tokamak building would be implemented on good bedrock made of limestone with a shear wave velocity of over 1,300 m/s. Such characteristics reduce the hazard when compared to sites with lower shear wave velocity where amplification of the seismic hazard could induce an increase of 30 percent. Four aspects will be discussed: The regulation: according to the classification of the facility, ITER will be a Nuclear Facility (Installation Nucléaire de Base); The implementation of this regulation for the proposed site. Description of site geology, tectonic, seismotectonic and consequently the set of design spectra and others characteristic of the site hazard (time history, length…); As an independent approach a probabilistic seismic hazard assessment of the site has been performed using a methodology which considers uncertainties (this work has been developed by the same team which had contributed to large parts of the GSHAP map study). Finally the fulfilment of the requirements and assumptions are discussed, according to IAEA guides for instance. As a conclusion of the studies the main characteristics of the Cadarache European site are discussed: as a reference the standardized 475 year mean return period hazard (i.e. 10 % of chance to occur in 50 years; 10,000 years return period in accordance with the ITER assumption; Frequency of the inspection level during construction and operation. Preliminary studies have shown that the European site proposal will ensure a low level of project risk with respect to the seismic hazard.


Corresponding Author:

Girard Jean-Philippe

Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-525 METHODOLOGY FOR REFERENCE ACCIDENTS DEFINITION FOR ITER

Pinna T., S. Raboin (1) J. Uzan-Elbez (1) N. Taylor (2) and EISS Team

(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Association Euratom UKAEA, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

The safety assessment of ITER presented in the Rapport Prémilinaire de Sûreté (RPrS) for the French Regulator imposes the definition and the study of a limited set of incidental and accidental sequences selected on a deterministic ground. Ultimate safety margins have been analysed through hypothetical sequences conservatively extrapolated from the more significant accidents. The exhaustiveness of the final list of accidents has been independently checked by application of two complementary probabilistic methods (top-down and bottom-up methods). The results of these complementary studies are available in the Generic Site Safety Report (GSSR), but they are not reproduced in the RPrS which needs to be only deterministic and homogenous to the approach required by the French regulator for reprocessing plants and laboratories. In the present state of the ITER project, the only parts concerned by the analysis are the tokamak, the tritium plant and the hot cells. The rationale for event selection consists first to identify every radiological source and its confinement barriers; failure of one or several of these barriers may then be presumed and a scenario defined, following a standardized grid; in any case, the calculations and analysis are achieved following a unique logical scheme to assure consistency and exhaustiveness of the report. Nine accident families have been defined: plasma events, loss of power events, in-vessel events, ex-vessel events, cryostat events, magnet events, maintenance events, tritium plant and fuel events, hot cells events. Calculations with qualified computer codes have shown that the consequences of any postulated accident respect the safety guidelines.


Corresponding Author:

Pinna T.

Association Euratom-ENEA, Via Enrico Fermi 27, 00044 Frascati (Roma), Italy

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-527 FIRE RISK ANALYSIS IN ITER TRITIUM BUILDING

Lignini Frank, (1) J.-Ph. Girard (2) L. Rodríguez-Rodrigo (2) J. Uzan-Elbez (3) M.-T. Porfiri and EISS Team

(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Asociación Euratom-CIEMAT, Avenida Complutense, 22, 28040 Madrid, Spain (3) Association Euratom-ENEA, Via Enrico Fermi 27, 00044 Frascati (Roma), Italy

Events as fire have been considered in ITER documentation of low probability and a general approach has been defined in [1] to be developed later for the ITER Specific site. It was said that “These hazards will be treated according to the industrial safety regulations and practices of the host country” [2]. In the framework of studies for the European ITER site in Cadarache, an assessment of fire hazard has been done in order to ensure compliance with French safety requirements. In this paper a summary of existing laws is presented and an example of the deterministic approach to be followed for the Preliminary Safety Report is given on the analysis of Tritium building design in [3]. For the study presented here the existing ITER design has been used. In a first step the analysis has been based on a screening by rooms where radioactive inventory is located. It should be demonstrated that fire would not induce severe consequences or that complementary design measures could be used for complying with the defined safety limits. In a second step a screening by important safety systems or components should be done, in order to demonstrate that they would not be destroyed by a fire. Rough calculations have been made showing the adequacy of the deterministic approach which is the most conservative one. Realistic calculations are being developed elsewhere. Lower values in the releases for realistic calculations than for the rough ones and the deterministic approach would be taken in support but not like a demonstration in the Preliminary Safety Report (RPrS). It is checked that the demonstration done is in agreement with ITER philosophy. The same analysis must be done for all the nuclear buildings. For the non-nuclear buildings it must be shown that a fire or any other event will not induce any radioactive releases to the environment higher than the dose limit to the population. [1] GSSR vol I-32 [2] FDR-PDD-5-Safety (FDR 2001) [3] Dossier d’Options de Sûreté – ITER, CEA/DSM/DRFC – Ind. C au 19/02/2002


Corresponding Author:

Lignini Frank

FRAMATOME-ANP, Division NOVATOME, 10-12 place Juliette Récamier 69006 Lyon, France

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-529 CHEMICAL RISK STUDIES INCLUDING BERYLLIUM AND CHEMICAL ZONING

Lebot P., J.Ph. Girard (1) J. Uzan-Elbez(1) and EISS Team

(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

The use of beryllium in a so called “Installation Nucléaire de Base, INB” which is the classification of ITER in France, with the foreseen quantities for ITER (13 t) is a new element in a licensing process in France. The aim of this work has been to gather the existing safety referential background in Europe and the world, in order to establish the basis for licensing ITER in Cadarache. This study has been developed in three parts: Synthesis of the international regulatory requirements concerning the use of beryllium; Identification of the generic design requirements associated in order to handle and minimise the beryllium risk; Preliminary beryllium risk analysis in order to make a proposition of the beryllium zoning for the Preliminary Safety Report (RPrS). From the point of view of regulatory requirements given toxicity limit values seem to be in good agreement with ITER guidelines. Concerning generic requirements some aspects to be taken into account in design phase have been identified and explained in detail, for example: fire in room with Be inventory decontamination procedure dismantling strategy for beryllium contamination level control confinement with specified leak per hour requirement on ventilation system (efficiency, pressure…) waste management cleaning worker training, dressing, medical monitoring special extinguisher materials for emergency situation Related to the preliminary beryllium risk analysis and zoning the details on following proposal will be presented: zoning for construction phase, start-up, tritium plasma test, operation, dismantling and non-controlled, controlled, breathing protection zone.


Corresponding Author:

Lebot P.

Hémisphère, 1 rue du château, 92200 Neuilly-sur-Seine

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-530 PROGRESS IN LICENSING ITER IN CADARACHE

Rodríguez-Rodrigo Lina, J.Ph. Girard (1) G. Marbach (2) J. Uzan-Elbez (1) P. Garin (2) and EISS Team

1 Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France 2 Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

In the framework of licensing ITER in France, a safety rapport called “Dossier d’Options de Sûreté” (DOS) has been submitted for examination by CEA, mandated to act on behalf of a future ITER organisation, to the French Safety authorities in March 2002. A thorough examination of the document has been performed by the technical body of the French Nuclear Safety Authority. The latter approved the contents of the DOS in November 2002 and issued recommendations to be taken into account in the next mandatory safety document to be prepared, the “Rapport Préliminaire de Sûreté”, RPrS. The RPrS is part of the application documentation for the Décret de Création, DAC, which will allow starting the construction of ITER. DAC must be submitted to a Public Enquiry. During 2002 and 2003 main elements for writing the RPrS have been elaborated and studies still to be done have been detected during this preparatory process. DARPE (“Demande d’Autorisation de Rejets d’Effluents et de Prélèvements d’Eau”) is an Authorisation Request for Effluents Release and Intake of Water. DARPE is the step following RPrS in the licensing process, but it is only mandatory for the operation and consequently, it could be deferred until then. Related safety documents concerning DARPE are mainly, a description of which kind of operation is planned, indicating the types of effluents to be processed, their origin, quantity, their radioactive and chemical composition, physical characteristics, treatment processes, waste release conditions and composition of the effluents to the rejected; a document indicating the effects of the operation (impact study). Non-radioactive effluents and industrial impact also have to be included (e.g. impact of a high voltage line on the countryside). Finally, the planned monitoring and intervention means in the case of incident or accident have to be presented. These documents will be extracted from RPrS file and rewritten in a public information style to be presented at the Public Enquiry. DAC and DARPE can be presented at the same time or separately to the Public Enquiry. The main content of these licensing documents is presented at the conference, as well as the main studies that have been launched and the schedule for the overall ITER licensing until start of operation.


Corresponding Author:

Rodríguez-Rodrigo Lina

3 Asociación Euratom-CIEMAT para Fusión, Avenida Complutense, 22, 28040 Madrid, Spain

- J - Power Plants, Safety and Environment, Socio-economics.

P4T-J-532 ALARA APPLIED TO ITER DESIGN. RADIOPROTECTION AND ZONING APPROACH

Uzan-Elbez J., J. Ph. Girard (1) L. Rodriguez (2) J.-M. Mure (3) M.-T. Porfiri(4) N. Taylor (5) and EISS Team

1 Direction de l’Énergie Nucléaire, CEA Cadarache, France 2 Asociación Euratom-CIEMAT, Madrid, Spain 3 Hémisphère, Neuilly-sur-Seine, France 4 Association Euratom-ENEA, Frascati, Italy 5 Association Euratom UKAEA, Abingdon, UK

Based on the existing data on ITER, a long list of references, and the safety options for licensing ITER in Cadarache, the objective of the present work is to demonstrate that ALARA principle has been implemented in the design of ITER and will be applied during ITER operation. ITER has established collective and individual dose objectives that meet regulatory limits and are in line with international trends and guidelines. The preliminary ORE estimate equals 250 H.mSv/y which is half the annual target. It essentially deals with maintenance activities. The main systems such as TCWS and in-vessel maintenance have been assessed. Assessment of ORE is still preliminary and needs to be completed, especially for systems like RH tools, NB cell, Hot Cell and radwaste building, whose design in still ongoing. Modelling assumptions should be confirmed and source term assessment consolidated together with maintenance programmes. Design alternatives such as remote handling of activated components (in-vessel and transfer to hot cell), radiation shielding of RH ports, delayed intervention after plasma burn, tritium plant design has been studied and compared as far as ORE is concerned; this leads to a choice of best available options as regards radiological exposure of workers. ITER maintenance and inspection procedures have been improved during ITER development and through the different steps of ITER design. Each design upgrade has been analysed bearing in mind the objective of dose reduction. Different examples of such an iterative process are given for different systems and show clear evidence of ALARA implementation. Analysis of preliminary results show that further optimisation is possible to ensure that radiological objectives are fulfilled and that doses in operation will be as low as reasonably achievable. It is concluded that main effort should be put on the demonstration that systems not assessed do not contribute to the total ORE, a continued Assessment of further modifications of the design; on the Analysis of uncertainties in order to check the margin set for collective dose (355 person.mSv/y); on the investigation of ways of improving the design of equipments that contribute mostly to the total ORE; on the improvement or set-up of maintenance and of inspection procedures in order to reduce work time or dose; and on the use of decision aiding tools for choosing among design alternatives (cost-benefit analysis for example)


Corresponding Author:

Uzan-Elbez J.

1 Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- K - Transfer of Technology.

P1C-K-131 100 KV SOLID-STATE SWITCH FOR FUSION HEATING SYSTEMS

Beaumont B. (3), R.Milly (1), H.Rigole(1) D.Chatroux (2) E.Bertrand(3), R.Brugnetti(3), F.Kazarian(3), M.Prou(3)

(1) A2E Enertronic, Département Puissance, 29 rue Condorcet, F-38090 VILLEFONTAINE, France (2) CEA, CEA-CENG, rue des Martyrs, 38054 GRENOBLE CEDEX 9, France (3) Association Euratom-CEA, CEA-Cadarache, 13108 St Paul-Lez-Durance , France

Power switching in RF heating systems is a delicate function as it is often linked to high power tube protection. In most RF systems, the end stage power tube is fed by a high voltage power supply (HVPS), which connection to the tube has to be interrupted in case of arc suspicion. The amount of energy that is allowable to be dissipated in the arc is in the range of 10 to 50 J, to limit the degradations observed of the tube structures. Furthermore, the HVPS is often common to several power tubes, and the loss of all the power from the group of tubes is to be avoided to minimize the perturbation on the plasma experiment. This protection function may be performed by ignitrons, or by using power switching on the HV side (pulse step modulator) in order to get a fast response, and the voltage regulation can be performed at the same time. Although these solutions seem attractive, They do not easily allow the sharing of a single HVPS between different power tubes. In the frame of the CIMES project, a new 700 kW CW klystron is being developed and will be installed in the existing transmitter. The present existing protection is based on spark gap triggered crowbar and is not compatible with the new tube which will be operated on varying cathode voltage between 50 and 75 kV to optimise the efficiency during long pulses. The existing protection shorts the HVPS to protect the tube, and the whole HVDC power supply is tripped, resulting in stopping the RF power from 4 klystrons at a time, with no recovery scheme during plasma shot. This new high power solid state switch uses MOSFETs matrix switches derived from a development in CEA for switching power supplies for copper vapor lasers in the frame of Uranium Vapor Laser Isotopic Separation (SILVA). 25 kV and 1600 A switches have been produced and have proved an exceptional reliability and tolerance to failure of discrete components. This technology has demonstrated to be cost effective, reliable and able to reach high performances. These techniques are used to design this new solid-state switch, which will be inserted in line just before the tube. It allows switching on and off the current between the HVPS and each klystron. The HVPS may be reconnected to the klystron after all tube parameters are back to normal, and the antenna can recover the power (and spectrum) during the plasma shot. This paper will detail the architecture of the switch, the technical specifications and the results of the prototype test program.


Corresponding Author:

Beaumont B. (3)

Association Euratom-CEA, CEA-Cadarache, 13108 St Paul-Lez-Durance ,France

- K - Transfer of Technology.

P1C-K-443 OVERVIEW OF CRYOGENIC REFRIGERATION SYSTEMS FOR THE THERMONUCLEAR FUSION

DAUGUET Pascale, BONNETON Michel ,DAUGUET Pascale, DELCAYRE Franck, HILBERT Benoit, RAVEX Alain

Advanced Technology Division, Air Liquide, 38360 Sassenage, France

Selected also for oral presentation O1B-K-443

Air Liquide DTA has largely participated for more than 20 years to the different stages of the fusion program supplying critical components to the main organizations such as JET, Tore Supra, Net Team, Iter, FZK, MFTF. The main contributions of Air Liquide are in the fields of * Large scale refrigeration of superconducting magnets (at 4 Kelvin or below). The 27-km-long Large Hadron Collider (LHC) of CERN will be briefly presented. It will make intensive use of superconducting magnets, operated below 2.0 K and will thus require for the first time on such a large scale high capacity refrigeration systems at temperatures ranging from 1.8 K to 80 K. Two refrigerators delivering 18 kW equivalent at 4.5 K were installed and commissioned in 2002. Making use of cryogenic centrifugal compressors in a series arrangement with room temperature screw compressors, the so-called “Cold Compression System” are able to absorb up to 2.4 kW at 1.8 K. * Ultra high vacuum by cryopumping. Should it be for pumping the divertor inside the plasma vessel, for the injection of pellets inside the torus or for plasma heating, a tokamak requests customised cryopumping systems. Development, design for construction, manufacturing, performance testing and commissioning will be addressed. * Gas purification and separation systems. Examples will be presented: hydrogen isotope separation obtained by cryogenic distillation at temperature around 20 K and atmosphere dryers and cold traps in nuclear environment. * Storage and distribution systems * Remote control of the equipments * Operation and maintenance of cryogenic and vacuum equipment.


Corresponding Author:

DAUGUET Pascale

Advanced Technology Division, Air Liquide, 38360 Sassenage, France

- K - Transfer of Technology.

P1C-K-499 DEVELOPMENT & APPLICATION OF MCNP AUTO-MODELING TOOL : MCAM 3.0

Xiaoping LIU, Luo YuetongCTong LiliCHuang QunyingCWu Yican and FDS Team

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031, P.R.China

MCNP is a transport simulation program using Monte Carlo technology widely used in design and evaluation of nuclear reactors. To achieve accurate results, preparation of detailed and fully described geometrical models is required, which is usually very time-consuming and error incidental by hand. MCAM (MCNP Auto-Modeling) is a Computer Aided Design (CAD) tool developed by the FDS Team (Institute of Plasma Physics, Chinese Academy of Science and Visualization & Cooperative Computing Division of Hefei University of Technology) for more than five years, it is designed to aid users of MCNP to prepare MCNPfs geometry model under visual environment. Main factual features of MCAM and application examples are presented. In MCAM, each cell of MCNPfs geometry is represented by a solid. Users can create solids in the following three ways: Users create solids directly in MCAM. MCAM provides various tools to support basic modeling function. To exchange data with other CAD software. MCAM can exchange data with other CAD software by STEP, IGES or SAT format file. To generate solids according to the existing MCNP geometry model. For MCNPfs geometry model, besides geometry information, some other information is also containedCsuch as materials and importance information, etc. In MCAM, users can input such information for given cells through dialog boxes. MCAM 3.0 has three main functions, they are: To convert CAD model into MCNP geometry model: For the given CAD model, MCAM can convert it into MCNP geometry. During this process, every solid in CAD model is considered as a cell in MCNP geometry model. At the same time, material No., material density, Neutron Importance and Photon Importance for each cell are also written into MCNP geometry model. To convert MCNP geometry model into CAD model: MCAM is originally conceived to convert CAD model into MCNP geometry model. To take use of lots of existing MCNP geometry model, MCAM 3.0 is extended to convert MCNP geometry into CAD model, either. Besides geometry information, material No., material density, neutron importance, photon importance are all remained. To maintain and modify CAD model: Many other functions are provided to make MCAM to be easily used, for example, exchanging data with other CAD systems, managing lots of solids of a CAD model, interface to specify material information for a given solid, and so on. All these functions make users to maintain and modify CAD model easily.


Corresponding Author:

Xiaoping LIU

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031, P.R.China



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