WorldWideScience

Sample records for replacing spent htr

  1. Maw and spent HTR Fuel Element Test storage in Boreholes in rock salt

    International Nuclear Information System (INIS)

    Barnert, E.; Brucher, P.H.; Kroth, K.; Merz, E.; Niephaus, D.

    1986-01-01

    The Budesminister fur Forschung und Technolgie (BMFT, Federal Ministry for Research and Technology) is sponsoring a project at the Kernforschungsanlage Julich (KFA, Juelich Nuclear Research Centre) entitled ''MAW and HTR Fuel Element Test disposal in Boreholes.'' The aim of this project is to develop a technique for the final disposal of (1) dissolver sludge, (2) cladding hulls/structural components and (3) spent HTR fuels elements in salt, and to test this technique in the abandoned Asse salt mine, including safety calculations and safety engineering demonstrations. The project is divided into the sub-projects I ''Disposal/sealing technique'' and II ''Retrievable disposal test.''

  2. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  3. PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

    International Nuclear Information System (INIS)

    De Wet, G. J.; Dent, C.

    2008-01-01

    A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

  4. A synthesis on the HTR scenario studies at CEA - HTR2008-58059

    International Nuclear Information System (INIS)

    Boucher, L.; Greneche, D.

    2008-01-01

    The aim of the studies is to assess the impact of the deployment of an HTR park replacing one part of the current PWR reactors. The other part of the current park is replaced by EPRs. In these scenarios, the annual electricity production is constant at 400 TWhe. This value corresponds roughly to the present nuclear electricity production in France. From 2002 to 2007, an important program study on HTR has been carried out by CEA and AREVA NC under the joint CEA - AREVA NC project 'prospective studies on the management of Plutonium and the back end of the cycle'. This program addresses core physic and scenario studies, and also the back end of the fuel cycle : reprocessing of spent fuel and HTR waste management. Some core physic studies have already been presented in the reference [1]. This paper presents the results of the scenario studies using two concepts: either the standard core of the Gas Turbine Modular Helium Reactor concept (GTMHR) with Uranium or Plutonium fuel, or the Multiple Fuel Rows Core (MFRC) dedicated to the actinide burning. The insertion of a new concept (fuel, reactor, process) must be evaluated in the global electronuclear system with an analysis of the impact on the fuel cycle (Enrichment, Fuel Fabrication, Reactor, Processing, Interim Storage, Waste storage). The scenario studies are used to evaluate different solutions to manage nuclear materials (uranium, plutonium) and wastes (minor actinides and fission products), from the present situation in France (closed cycle with storage of used MOX fuels) until the final equilibrium: mixed nuclear park with EPR and HTR. These studies allow to calculate material flows and inventories of these elements in each step of the fuel cycle. The simulation of transient scenarios from the present situation to the future situation is performed with the COSI code. HTR reactors feature a high flexibility with regard to fuel cycle options. Several versions of core have been investigated, with different type of

  5. A real-time data acquisition and processing system for the analytical laboratory automation of a HTR spent fuel reprocessing facility

    International Nuclear Information System (INIS)

    Watzlawik, K.H.

    1979-12-01

    A real-time data acquisition and processing system for the analytical laboratory of an experimental HTR spent fuel reprocessing facility is presented. The on-line open-loop system combines in-line and off-line analytical measurement procedures including data acquisition and evaluation as well as analytical laboratory organisation under the control of a computer-supported laboratory automation system. In-line measurements are performed for density, volume and temperature in process tanks and registration of samples for off-line measurements. Off-line computer-coupled experiments are potentiometric titration, gas chromatography and X-ray fluorescence analysis. Organisational sections like sample registration, magazining, distribution and identification, multiple data assignment and especially calibrations of analytical devices are performed by the data processing system. (orig.) [de

  6. Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

    International Nuclear Information System (INIS)

    Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

    2008-01-01

    South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

  7. The Renewal of HTR Development in Europe

    International Nuclear Information System (INIS)

    Hittner, Dominique

    2002-01-01

    The European HTR-Technology Network (HTR-TN), created in 2000, presently groups 20 organisations from European nuclear research and industry for developing the technologies of direct-cycle modular HTRs, which presently raise a large world-wide interest, because of their high potential for economic competitiveness, natural resource sparing, safety and minimisation of the waste impacts, in line with the goals of sustainable development of Generation IV. All aspects of HTR technologies are addressed by HTR-TN, from the reactor physics to the development of materials, fuel and components. Most of this activity is supported by the European Commission in the frame of its 5. EURATOM Framework Programme. The first results of HTR-TN programme are given: the analysis of the reactor physics international benchmark on the commissioning tests of HTTR (Japan), the long term behaviour of spent HTR fuel in geologic disposal conditions, the preparation of a very high burnup fuel irradiation and the development of fabrication processes for producing high performance coated particles, etc. (authors)

  8. HTR-TN a European network for the development of HTR technology

    International Nuclear Information System (INIS)

    Von Lensa, W.

    2001-01-01

    A network called High-temperature reactor technology network (HTR-TN) has been created at a European level to coordinate works and knowledge on the subject with a long-term perspective and to serve as a channel for international collaboration. An analysis confirmed that the obvious economic penalty of HTR due to its low density power could be compensated by the combination of recent advances that may completely change the positioning of HTR on the energy market: -) the modular concept allowed to get a reactor free from core melt risk without intervention of any active safety system, implying a drastic simplification of the design of the reactor and the safety systems as well as a standardisation and potential for shop fabrication in series; -) the development of gas turbines, the efficiency of which increased, in 10 years, from 35% till 50% and more, enabling to consider suppression of the secondary system; -) the ultra high burn-up potential of HTR fuel and the possibility for direct disposal of spent HTR fuel elements that may reduce cost of the fuel cycle and contribute to the reduction of civil and military plutonium stockpiles. (A.C.)

  9. HTR Development in China

    International Nuclear Information System (INIS)

    Wang Dazhong

    2014-01-01

    The roles of HTRs in China: 1. Due to the inherent safety features, high efficiency of electricity generation, site flexibility, the modular HTR can act as a supplement to LWR for small and medium size power generation. 2. Co-generation to supply steam up to 600℃, for petroleum refinery, oil sand and oil shale processing, sea water desalination and district heating, etc. 3. Hydrogen production at 900~1000 ℃ by V/HTR. Conclusions and prospects: • China’s energy system will experience transition and reform in the future; • Nuclear energy will play an irreplaceable role in China’s energy development; • Due to the excellent features of inherent safety, the HTR is a promising technology for electricity generation and process heat utilization; • Further international cooperation and exchanges need to be enhanced

  10. Spent Mushroom Waste as a Media Replacement for Peat Moss in Kai-Lan (Brassica oleracea var. Alboglabra Production

    Directory of Open Access Journals (Sweden)

    H. Sendi

    2013-01-01

    Full Text Available Peat moss (PM is the most widely used growing substrate for the pot culture. Due to diminishing availability and increasing price of PM, researchers are looking for viable alternatives for peat as a growth media component for potted plants. A pot study was conducted with a view to investigate the possibility of using spent mushroom waste (SMW for Kai-lan (Brassica oleracea var. Alboglabra production replacing peat moss (PM in growth media. The treatments evaluated were 100% PM (control, 100% SMW, and mixtures of SMW and PM in different ratios like 1 : 1, 1 : 2, and 2 : 1 (v/v with/without NPK amendment. The experiment was arranged in a completely randomized design with five replications per treatment. Chemical properties like pH and salinity level (EC of SMW were within the acceptable range of crop production but, nutrient content, especially nitrogen content was not enough to provide sufficient nutrition to plant for normal growth. Only PM (100% and SMW and PM mixture in 1 : 1 ratio with NPK amendment performed equally in terms of Kai-lan growth. This study confirms the feasibility of replacing PM by SMW up to a maximum of 50% in the growth media and suggests that NPK supplementation from inorganic sources is to ensure a higher productivity of Kai-lan.

  11. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

  12. Tritium in HTR systems

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-07-01

    Starting from the basis of the radiological properties of tritium, the provisions of present-day radiation protection legislation are discussed in the context of the handling of this radionuclide in HTR plants. Tritium transportation is then followed through from the place of its creation up until the sink, i.e. disposal and/or environmental route, and empirical values obtained in experiments and in plant operation translated into guidelines for plant design and planning. The use of the example of modular HTR plants permits indication that environmental contamination via the 'classical' routes of air and water emissions, and contamination of products, and resulting consumer exposure, are extremely low even on the assumption of extreme conditions. This leads finally to a requirement that the expenditure for implementation of measures for further reduction of tritium activity rates be measured against low radiological effect. (orig.) [de

  13. International HTR activities

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    Asea Brown Boveri AG (ABB) and their subsidiary High Temperature Reactor Construction GmbH (HRB) have brought the pebble bed high temperature reactor to the edge of being ready for the market with the construction and operation of the AVR reactor at Juelich and the THTR 300 at Hamm-Uentrop. Siemens/Interatom have developed the HTR modular concept and, together with their partners HRB, KFA, Rheinbraun Bergbauforschung have taken the nuclear process heat project to its present advanced state of development. The further introduction of the HTR to the market is a long-term objective, due to the present market situation. ABB and Siemens AG have therefore agreed to collaborate by forming a joint company. (orig.)

  14. Future Development of Modular HTGR in China after HTR-PM

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wang, Haitao; Dong Yujie; Li Fu

    2014-01-01

    The modular high temperature gas-cooled reactor (MHTGR) is an inherently safe nuclear energy technology for efficient electricity generation and process heat applications. The MHTGR is promising in China as it may replace fossil fuels in broader energy markets. In line with China’s long-term development plan of nuclear power, the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University developed and designed a MHTGR demonstration plant, named high-temperature gas-cooled reactor-pebble bed module (HTR-PM). The HTR-PM came into the construction phase at the end of 2012. The HTR-PM aims to demonstrate safety, economic potential and modularization technologies towards future commercial applications. Based on experiences obtained from the HTR-PM project with respect to design, manufacture, construction, licensing and project management, a further step aiming to promote commercialization and market applications of the MHTGR is expected. To this purpose, INET is developing a commercialized MHTGR named HTR-PM600 and a conceptual design is under way accordingly. HTR-PM600 is a pebble-bed MHTGR power generation unit with a six-pack of 250MWth reactor modules. The objective is to cogenerate electricity and process heat flexibly and economically in order to meet a variety of market needs. The design of HTR-PM600 closely follows HTR-PM with respect to safety features, system configuration and plant layout. HTR-PM600 has the six modules feeding one steam turbine to generate electricity with capacity to extract high temperature steam from various interfaces of the turbine for further process heat applications. A standard plant consists of two HTR-PM600 units. Based on the economic information of HTR-PM, a preliminary study is carried out on the economic prospect of HTR-PM600. (author)

  15. HTR characteristics affecting reactor physics

    International Nuclear Information System (INIS)

    Ehlers, K.

    1980-01-01

    A physical description of high-temperature has-cooled reactors is given, followed by an overview of HTR characteristics. The emphasis is placed on the HTR fuel cycle alternatives and thermohydraulics of pebble bed core. Some prospects of HTRs in the Federal Republic of Germany are also presented

  16. Symbiosis of near breeder HTR's with hybrid fusion reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1978-07-01

    In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

  17. HTR Plans in Poland

    International Nuclear Information System (INIS)

    Sobolewski, Józef

    2017-01-01

    Target for HTR: Polish Heat Market: Today 100% heat market is dominated by fossil fuels; mostly coal in district heating and coal and gas in industry heat generation. Huge potential for nuclear reactors Currently can be addressed only in terms of LWR, i.e. T <250 ° C, useful in district heating, but not in industry. Need for new technologies •HTGR (High Temperature Gas Reactor) ~600°C, e.g. for industry steam generation. •VHTR (Very High Temperature Reactor), ... ~1000°C, e.g. for hydrogen production

  18. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  19. Experimental study of optimal self compacting concrete with spent foundry sand as partial replacement for M-sand using Taguchi approach

    Directory of Open Access Journals (Sweden)

    Nirmala D.B.

    2016-06-01

    Full Text Available This paper presents the application of Taguchi approach to obtain optimal mix proportion for Self Compacting Concrete (SCC containing spent foundry sand and M-sand. Spent foundry sand is used as a partial replacement for M-sand. The SCC mix has seven control factors namely, Coarse aggregate, M-sand with Spent Foundry sand, Cement, Fly ash, Water, Super plasticizer and Viscosity modifying agent. Modified Nan Su method is used to proportion the initial SCC mix. L18 (21×37 Orthogonal Arrays (OA with the seven control factors having 3 levels is used in Taguchi approach which resulted in 18 SCC mix proportions. All mixtures are extensively tested both in fresh and hardened states to verify whether they meet the practical and technical requirements of SCC. The quality characteristics considering “Nominal the better” situation is applied to the test results to arrive at the optimal SCC mix proportion. Test results indicate that the optimal mix satisfies the requirements of fresh and hardened properties of SCC. The study reveals the feasibility of using spent foundry sand as a partial replacement of M-sand in SCC and also that Taguchi method is a reliable tool to arrive at optimal mix proportion of SCC.

  20. Experimental study of optimal self compacting concrete with spent foundry sand as partial replacement for M-sand using Taguchi approach

    Science.gov (United States)

    Nirmala, D. B.; Raviraj, S.

    2016-06-01

    This paper presents the application of Taguchi approach to obtain optimal mix proportion for Self Compacting Concrete (SCC) containing spent foundry sand and M-sand. Spent foundry sand is used as a partial replacement for M-sand. The SCC mix has seven control factors namely, Coarse aggregate, M-sand with Spent Foundry sand, Cement, Fly ash, Water, Super plasticizer and Viscosity modifying agent. Modified Nan Su method is used to proportion the initial SCC mix. L18 (21×37) Orthogonal Arrays (OA) with the seven control factors having 3 levels is used in Taguchi approach which resulted in 18 SCC mix proportions. All mixtures are extensively tested both in fresh and hardened states to verify whether they meet the practical and technical requirements of SCC. The quality characteristics considering "Nominal the better" situation is applied to the test results to arrive at the optimal SCC mix proportion. Test results indicate that the optimal mix satisfies the requirements of fresh and hardened properties of SCC. The study reveals the feasibility of using spent foundry sand as a partial replacement of M-sand in SCC and also that Taguchi method is a reliable tool to arrive at optimal mix proportion of SCC.

  1. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

  2. French programme for HTR fuel

    International Nuclear Information System (INIS)

    Gillet, R.M.

    1991-01-01

    It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

  3. HTR-10 management information system

    International Nuclear Information System (INIS)

    Liu Ruoxiao; Wu Zhongwang; Xi Shuren

    2000-01-01

    The HTR-10 Management information system (REMIS) strengthens the managerial level and usage of the information of HTR-10, thereby enhances the ability and efficiency of the design and management work. REMIS is designed based on the Client/Server framework. Database management system is SQL Server 6.5 for NT, While the client side is developed by Borland C ++ Builder, and it is based on Windows 95/98. The network protocol is TCP/IP. REMIS collects date of the HTR-10 at four parameters: Reactor properties, Design parameters, Equipment properties Reactor system flow charts. Final discussing extended prospect of REMIS

  4. Replacement

    Directory of Open Access Journals (Sweden)

    S. Radhakrishnan

    2014-03-01

    Full Text Available The fishmeal replaced with Spirulina platensis, Chlorella vulgaris and Azolla pinnata and the formulated diet fed to Macrobrachium rosenbergii postlarvae to assess the enhancement ability of non-enzymatic antioxidants (vitamin C and E, enzymatic antioxidants (superoxide dismutase (SOD and catalase (CAT and lipid peroxidation (LPx were analysed. In the present study, the S. platensis, C. vulgaris and A. pinnata inclusion diet fed groups had significant (P < 0.05 improvement in the levels of vitamins C and E in the hepatopancreas and muscle tissue. Among all the diets, the replacement materials in 50% incorporated feed fed groups showed better performance when compared with the control group in non-enzymatic antioxidant activity. The 50% fishmeal replacement (best performance diet fed groups taken for enzymatic antioxidant study, in SOD, CAT and LPx showed no significant increases when compared with the control group. Hence, the present results revealed that the formulated feed enhanced the vitamins C and E, the result of decreased level of enzymatic antioxidants (SOD, CAT and LPx revealed that these feeds are non-toxic and do not produce any stress to postlarvae. These ingredients can be used as an alternative protein source for sustainable Macrobrachium culture.

  5. Clinical Validation of Therapeutic Drug Monitoring of Imipenem in Spent Effluent in Critically Ill Patients Receiving Continuous Renal Replacement Therapy: A Pilot Study.

    Science.gov (United States)

    Wen, Aiping; Li, Zhe; Yu, Junxian; Li, Ren; Cheng, Sheng; Duan, Meili; Bai, Jing

    2016-01-01

    The primary objective of this pilot study was to investigate whether the therapeutic drug monitoring of imipenem could be performed with spent effluent instead of blood sampling collected from critically ill patients under continuous renal replacement therapy. A prospective open-label study was conducted in a real clinical setting. Both blood and effluent samples were collected pairwise before imipenem administration and 0.5, 1, 1.5, 2, 3, 4, 6, and 8 h after imipenem administration. Plasma and effluent imipenem concentrations were determined by reversed-phase high-performance liquid chromatography with ultraviolet detection. Pharmacokinetic and pharmacodynamic parameters of blood and effluent samples were calculated. Eighty-three paired plasma and effluent samples were obtained from 10 patients. The Pearson correlation coefficient of the imipenem concentrations in plasma and effluent was 0.950 (Pimipenem concentration ratio was 1.044 (95% confidence interval, 0.975 to 1.114) with Bland-Altman analysis. No statistically significant difference was found in the pharmacokinetic and pharmacodynamic parameters tested in paired plasma and effluent samples with Wilcoxon test. Spent effluent of continuous renal replacement therapy could be used for therapeutic drug monitoring of imipenem instead of blood sampling in critically ill patients.

  6. Worldwide status of HTR development

    International Nuclear Information System (INIS)

    1978-06-01

    The International Atomic Energy Agency convened a technical committee meeting on high temperature reactors (HTRs) from 12-14 Dec. 1977 at Agency Headquarters to provide a forum for the exchange of information on the status of HTR development programmes and to receive advice on the Agency programme in this field. The continuing high level of international interest in HTRs was evidenced by the participation from 11 countries and 2 organizations: Austria, Belgium, France, Federal Republic of Germany, Japan, Netherlands, Poland, Switzerland, Union of Soviet Socialist Republics, United Kingdom of Great Britain, United States of America, Commission of the European Communities, and the OECD Nuclear Energy Agency. In order to promote the continuing exchange of technical information through the offices of the IAEA, a recommendation was made that the Agency establish a standing International Working Group on High Temperature Reactors (IWGHTR). This recommendation is being implemented in 1978. Considerable information on recent progress in HTR development was present at the technical committee meeting in technical reports and in progress reports on HTR development programmes. Since this material will not be published, this summary report on the worldwide status of HTR development at the beginning of 1978 has been prepared, based primarily on information presented at the December 1977 meeting

  7. Progress of the HTR-10 project

    International Nuclear Information System (INIS)

    Zhong, D.; Xu, Y.

    1996-01-01

    This paper briefly introduces the main technical features and the design specifications of the HTR-10. Present status and main progress of the license applications, the design and manufacture of the main components and the engineering experiments as well as the construction of the HTR-10 are summarized. (author). 3 tabs

  8. Report on the ANSTO application for a licence to construct a Replacement Research Reactor, addressing seismic analysis and seismic design accident analysis, spent fuel and radioactive wastes

    International Nuclear Information System (INIS)

    2002-02-01

    The Report of the Nuclear Safety Committee (NSC) covers specific terms of reference as requested by the Chief Executive Officer of ARPANSA. The primary issue for the Working Group(WG) consideration was whether ANSTO had demonstrated: (i) that the overall approach to seismic analysis and its implementation in the design is both conservative and consistent with the international best practice; (ii) whether the full accident analysis in the Probabilistic Safety Assesment Report (PSAR) satisfies the radiation dose/frequency criteria specified in ARPANSA's regulatory assessment principle 28 and the assumptions used in the Reference Accident for the siting assessment have been accounted for in the PSAR; and (iii) the adequacy of the strategies for managing the spent fuel as proposed to be used in the Replacement Research Reactor and other radioactive waste (including emissions, taking into account the ALARA criterion) arising from the operation of the proposed replacement reactor and radioisotope production. The report includes a series of questions that were asked of the Applicant in the course of working group deliberations, to illustrate the breadth of inquiries that were made. The Committee noted that replies to some questions remain outstanding at the date of this document. The NSC makes a number of recommendations that appear in each section of the document, which has been compiled in three parts representing the work of each group. The NSC notes some lack of clarity in what was needed to be considered at this approval stage of the project, as against information that would be required at a later stage. While not in the original work plan, recent events of September 11, 2001 also necessitated some exploration of issues relating to construction security. Copyright (2002) Commonwealth of Australia

  9. Effects of Replacing Pork Back Fat with Canola and Flaxseed Oils on Physicochemical Properties of Emulsion Sausages from Spent Layer Meat

    Directory of Open Access Journals (Sweden)

    Ki Ho Baek

    2016-06-01

    Full Text Available The objective of this study was to investigate the effects of canola and flaxseed oils on the physicochemical properties and sensory quality of emulsion-type sausage made from spent layer meat. Three types of sausage were manufactured with different fat sources: 20% pork back fat (CON, 20% canola oil (CA and 20% flaxseed oil (FL. The pH value of the CA was significantly higher than the others (p<0.05. The highest water holding capacity was also presented for CA; in other words, CA demonstrated a significantly lower water loss value among the treatments (p<0.05. CA had the highest lightness value (p<0.05. However, FL showed the highest yellowness value (p<0.05 because of its own high-density yellow color. The texture profile of the treatments manufactured with vegetable oils showed higher values than for the CON (p<0.05; furthermore, CA had the highest texture profile values (p<0.05 among the treatments. The replacement of pork back fat with canola and flaxseed oils in sausages significantly increased the omega-3 fatty acid content (p<0.05 over 15 to 86 times, respectively. All emulsion sausages containing vegetable oil exhibited significantly lower values for saturated fatty acid content and the omega-6 to omega-3 ratios compared to CON (p<0.05. The results show that using canola or flaxseed oils as a pork fat replacer has a high potential to produce healthier products, and notably, the use of canola oil produced characteristics of great emulsion stability and sensory quality.

  10. Effects of Replacing Pork Back Fat with Canola and Flaxseed Oils on Physicochemical Properties of Emulsion Sausages from Spent Layer Meat.

    Science.gov (United States)

    Baek, Ki Ho; Utama, Dicky Tri; Lee, Seung Gyu; An, Byoung Ki; Lee, Sung Ki

    2016-06-01

    The objective of this study was to investigate the effects of canola and flaxseed oils on the physicochemical properties and sensory quality of emulsion-type sausage made from spent layer meat. Three types of sausage were manufactured with different fat sources: 20% pork back fat (CON), 20% canola oil (CA) and 20% flaxseed oil (FL). The pH value of the CA was significantly higher than the others (p<0.05). The highest water holding capacity was also presented for CA; in other words, CA demonstrated a significantly lower water loss value among the treatments (p<0.05). CA had the highest lightness value (p<0.05). However, FL showed the highest yellowness value (p<0.05) because of its own high-density yellow color. The texture profile of the treatments manufactured with vegetable oils showed higher values than for the CON (p<0.05); furthermore, CA had the highest texture profile values (p<0.05) among the treatments. The replacement of pork back fat with canola and flaxseed oils in sausages significantly increased the omega-3 fatty acid content (p<0.05) over 15 to 86 times, respectively. All emulsion sausages containing vegetable oil exhibited significantly lower values for saturated fatty acid content and the omega-6 to omega-3 ratios compared to CON (p<0.05). The results show that using canola or flaxseed oils as a pork fat replacer has a high potential to produce healthier products, and notably, the use of canola oil produced characteristics of great emulsion stability and sensory quality.

  11. Characteristic analysis of rotor dynamics and experiments of active magnetic bearing for HTR-10GT

    International Nuclear Information System (INIS)

    Yang Guojun; Xu Yang; Shi Zhengang; Gu Huidong

    2005-01-01

    A 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University of China. The helium turbine and generator system of 10 MW high temperature gas-cooled reactor (HTR-10GT) is the second phase for the HTR-10 project. It is to set up a direct helium cycle to replace the current steam cycle. The active magnetic bearing (AMB) instead of ordinary mechanical bearing was chosen to support the rotor in the HTR-10GT. This rotor is vertically mounted to hold the turbine machine, compressors and the power generator together. The rotor's length is 7 m, its weight is about 1500 kg and the rotating speed is 15000 r/min. The structure of the rotor is so complicated that dynamic analysis of the rotor becomes difficult. One of the challenging problems is to exceed natural frequencies with enough stability and safety during reactor start up, power change and shutdown. The dynamic analysis of the rotor is the base for the design of control system. It is important for the rotor to exceed critical speeds. Some kinds of software and methods, such as MSC.Marc, Ansys, and the Transfer Matrix Method, are compared to fully analyze rotor dynamics characteristic in this paper. The modal analysis has been done for the HTR-10GT rotor. MSC.Marc was finally selected to analyze the vibration mode and the natural frequency of the rotor. The effects of AMB stiffness on the critical speeds of the rotor were studied. The design characteristics of the AMB control system for the HTR-10GT were studied and the related experiment to exceed natural frequencies was introduced. The experimental results demonstrate the system functions and validate the control scheme, which will be used in the HTR-10GT project. (authors)

  12. Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

    Directory of Open Access Journals (Sweden)

    Ortego Pedro

    2017-01-01

    Full Text Available In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

  13. Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

    Science.gov (United States)

    Ortego, Pedro; Rodriguez, Alain; Töre, Candan; Compadre, José Luis de Diego; Quesada, Baltasar Rodriguez; Moreno, Raul Orive

    2017-09-01

    In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S) developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

  14. Effect of replacing dietary vitamin E by sage on performance and meatiness of spent hens, and the oxidative stability of sausages produced from their meat.

    Science.gov (United States)

    Loetscher, Y; Kreuzer, M; Albiker, D; Stephan, R; Messikommer, R E

    2014-01-01

    A total of 3960 hens (half ISA Warren and half Dekalb White) were housed in 18 compartments with 220 hens each. The effect of replacing dietary vitamin E by sage on productivity, meat yield and oxidative stability of sausages was studied. One third of all animals received either a vitamin E deficient diet (negative control) or diets supplemented with 30 mg/kg α-tocopherylacetate (positive control) or 25 g sage leaves/kg. At slaughter, meat yield was assessed and sausages were produced (n = 12 per treatment). The omission of vitamin E did not impair the oxidative stability of the raw sausage material or the spiced sausages in comparison to the positive control. Sage supplementation improved oxidative stability after 7 m of frozen storage, but not after 1, 4 and 10 m. Spice addition during meat processing had an antioxidant effect regardless of dietary treatment. Diet supplementation of any type did not affect laying performance and sausage meat yield. Feeding antioxidants to spent hens seemed to be not as efficient as in growing chickens, while seasoning with spices during sausage production proved to be a feasible way to delay lipid oxidation.

  15. Status of the HTR 500 design program

    International Nuclear Information System (INIS)

    Baust, E.; Arndt, E.

    1987-01-01

    Since 1982 BBC/HRB have offered the HTR 500 as the follow-on project of the THTR 300, the first large pebble bed reactor. The technical concept of the HTR-500 largely corresponds to the THTR 300 which has been in operation for almost 2 years now. In developing the design concept of the HTR 500 the ideas and demands of the reactor users in the FRG interested in the HTR have been taken into consideration to a large extent. In 1982 these potential users formed a working group 'Arbeitsgemeinschaft Hochtemperaturreaktor' (AHR), representing 16 power indusry companies and in early 1983, awarded a contract to HRB to perform a conceptual design study on the HTR 500. Within this conceptual design study BBC/HRB developed the safety concept of the HTR 500, prepared a detailed description of the overall power plant, and performed a cost calculation. These activities were completed in 1984. Based on the positive results of this conceptual design study, BBC/HRB are expecting to be granted a design contract by the users company Hochtemperaturreaktor GmbH (HRG) to establish the final complete design plans and documents for the HTR 500. (author)

  16. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  17. For a Global HTR Marketing Initiative

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Venneri, Francesco; Richards, Matthew

    2014-01-01

    HTRs are at a crossroads in their history. The technology is proven and the current technical developments relatively mastered but the marketing track record is disappointing. This paper comes to the conclusion that an international, collaborative marketing and communication plan must be implemented in order to address the marketing bottleneck of HTRs. The paper reflects about the HTR product specificities, its unique selling points and its positioning against other nuclear designs and gas cogeneration. It summarises the global market status and demonstrates that the global market for HTRs is there, for electricity generation, industrial cogeneration and polygeneration. The paper finally argues that HTR vendors have a shared interest to unite in order to succeed in activating the market demand for HTR, and suggests an action plan for an international collaboration among HTR vendors to market and communicate globally on HTRs and reach together a critical mass of business leads worldwide, a mutually beneficial outcome. (author)

  18. Notes on HTR applications in methanol production

    International Nuclear Information System (INIS)

    Santoso, B.; Barnert, H.

    1997-01-01

    Notes on the study of HTR applications are presented. The study in particular should be directed toward the most feasible applications of HTR for process heat generation. A prospective study is the conversion of CO 2 gas from Natuna to methanol or formic acid. Further studies needs to be deepened under the auspices of IAEA and countries that have similar interest. (author). 3 refs, 3 figs

  19. KWU's modular approach to HTR commercialization

    International Nuclear Information System (INIS)

    Frewer, H.; Weisbrodt, I.

    1983-01-01

    As a way of avoiding the uncertainties, delays and unacceptable commercial risks which have plagued advanced reactor projects in Germany, KWU is advocating a modular approach to commercialization of the high-temperature reactor (HTR), using small size standard reactor units. KWU has received a contract for the study of a co-generation plant based on this modular system. Features of the KWU modular HTR, process heat, gasification, costs and future development are discussed. (UK)

  20. Present status of research and development for HTR in China

    Energy Technology Data Exchange (ETDEWEB)

    Dazhong, Wang; Daxin, Zhong; Yuanhul, Xu [Institute of Nuclear Energy Technology, Tsinghua University, Beijing (China)

    1990-07-01

    The HTR R and D Project is being carried out in the relevant institutions in China. Some topics are covered such as, fuel element technology, graphite development, fuel element handling system, helium technology, fuel reprocessing technology as well as HTR design study. Some results of HTR research work are described. In addition, to provide a test facility for investigation of HTR Module reactor safety and process heat application of HTR, a joint project on building a 10 MW test HTR with Siemens-Interatom, KFA Juelich and INET is going on. The conceptual design of 10 MW test HTR has been completed by the joint group. In parallel the application study of HTR Module is being carried out for the oil industry, petrochemical industry as well as power generation. Some preliminary results of the application study, for example, for heavy oil recovery on Shengli oil field and process heat application in Yan shan petroleum company, are described. (author)

  1. The HTR-PM Plant Full Scope Training Simulator

    International Nuclear Information System (INIS)

    Wang Junsan; Wang Yuding; Zhou Shuyong; Cai Ruizhong; Cao Jianting

    2014-01-01

    This paper describes the technical aspects of the Full Scope Training Simulator developed for HTR-PM Plant in Shidao Bay, Shandong Province, China. An overview of the HTR-PM plant and simulator structure is presented. The models developed for the simulator are discussed in detail. Some important verification tests have been conducted on the HTR-PM Plant Training Simulator. (author)

  2. The HTR safety concept demonstrated by selected examples

    International Nuclear Information System (INIS)

    Sommer, H.; Stoelzl, D.

    1981-01-01

    The licensing experience gained in the Federal Republic of Germany is based on the licensing procedures for the THTR-300 and the HTR-1160. In the course of the licensing procedures for these reactors a safety concept for an HTR has been developed. This experience constitutes the basis for the design of future HTR's. (author)

  3. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  4. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  5. Postirradiation examination of HTR fuel

    International Nuclear Information System (INIS)

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

  6. MCNP qualification on the HTR critical configurations: HTTR, HTR10 and PROTEUS results

    Energy Technology Data Exchange (ETDEWEB)

    TRAKAS, Christos; STOVEN, Gilles [AREVA NP, Tour Areva, 92084 Paris La Defence Cedex (France)

    2008-07-01

    Recent critical experiments, including PROTEUS, HTTR and HTR-10 provide a reliable qualification base for HTR criticality predictions. The fuel tested in these experiments, be it hexagonal block or pebble type, is irradiated in a spectrum comparable to that of the HTR planned by AREVA NP. The neutron spectrum is comparable in all three cases; the mean C/M value for all critical cases is less than +350 pcm (JEF2.2), +250 pcm (JEFF3.1) and +60 pcm (ENDF BVI). The C/M obtained for the rods worth, the reaction rates and the isothermal coefficient are very satisfactory. (authors)

  7. Capital costs of modular HTR reactors

    International Nuclear Information System (INIS)

    Kugeler, K.; Froehling, W.

    1993-01-01

    A decisive factor in the introduction of a reactor line, in addition of its safety, which should exclude releases of radioactivity into the environment, is its economic development and, consequently, its competitiveness. The costs of the pressurized water reactor are used for comparison with the modular HTR reactor. If the measures proposed for evolutionary increases in safety of the PWR are taken, cost increases will have to be expected for that line. The modular HTR can now attain specific construction costs of 3000 deutschmarks per electric kilowatt. Mass production and the introduction of cost-reducing innovations can improve the economy of this line even further. In this way, the modular HTR concept offers the possibility to vendors and operators to set up new economic yardsticks in safety technology. (orig.) [de

  8. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    International Nuclear Information System (INIS)

    Bomboni, Eleonora

    2008-01-01

    The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

  9. HTR fuel development for advanced application

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.; Graham, L.W.; Hick, H.

    1975-01-01

    The advantages of the HTR for nuclear steam supply systems are briefly outlined. Due to its great design flexibility a number of different designs have evolved and the main characteristics of existing experimental prototype and power reactor HTR designs are summarized. The present state of coated particle fuel, particularly with regard to performance, is considered. Some implications of producing higher temperatures are discussed. Finally some of the developments in progress such as minimising the temperature drop between fuel and coolant, and of improving fuel performance by better fission product retention, better chemical stability, and the use of alternative coated materials, are discussed. (U.K.)

  10. Actinide production in different HTR-fuel cycle concepts

    International Nuclear Information System (INIS)

    Filges, D.; Hecker, R.; Mirza, N.; Rueckert, M.

    1978-01-01

    At the 'Institut fuer Reaktorentwicklung der Kernforschungsanlage Juelich' the production of α-activities in the following HTR-OTTO cycle concepts were studied: 1. standard HTR cycle (U-Th); 2. low enriched HTR cycle (U-Pu); 3. near breeder HTR cycle (U-Th); 4. combined system (conventional and near breeder HTR). The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The calculations were performed by the short depletion code KASCO and the well-known ORIGEN program

  11. HTR fuel research in the HTR-TN network on the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M. [HFR Unit, Institute for Advanced Materials, European Commission, Joint Research Centre, Petten (Netherlands); Languille, A. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Guillermier, P. [FRAMATOME ANP, 69 - Lyon (France); Bakker, K. [Nuclear Research and Consultancy Group, Petten (Netherlands); Nabielek, H. [Forschungszentrum Juelich (Germany)

    2001-07-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  12. HTR fuel research in the HTR-TN network on the high flux reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M.; Languille, A.; Guillermier, P.; Bakker, K.; Nabielek, H.

    2001-01-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  13. Fuel cycle studies for the Dragon HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A; Nunn, R M; Twitchin, A E

    1971-02-15

    This note reports the progress made at B.N.L. in the study of the fuel cycle for the HTR design described by Daub (1970). The primary purpose of the study is to examine the special problems of the approach to equilibrium fuel cycle.

  14. The physics design of the HTR-1160

    International Nuclear Information System (INIS)

    Huebner, A.; Brandes, S.

    1975-01-01

    This paper describes the physica design of the reactor core of the helium cooled, graphite moderated high-temperature reactor HTR-1160. A discussion is given of the design criteria, the calculational methods, the burnup cycle, the power distribution and the reactivity control. (orig.) [de

  15. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  16. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  17. Status of development of the HTR module

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1989-01-01

    Growing concern about the rising global temperature of the earth due to the ''Greenhouse Effect'' is increasingly focussing worldwide interest on passively safe reactors for heat and power production. In this context the development status of the HTR-Module designed by the Siemens-Group merits strong interest. The HTR-Module has a high degree of passive safety features. Even in case of hypothetical accidents the decay heat is dissipated from the primary system to the environment by passive measures alone i.e. by heat conduction, convection and radiation. The detailed engineering for the HTR-Module continues to progress. In addition to the engineering for the layout considerable progress has been made in the detailed engineering for specific components - e.g. pressure vessel, steam generator, hot gas duct, blower etc. - and specific systems - e.g. first core, helium purification system, reactor safety system, reactor control etc. The procedure for the conceptual licence has been continued. A large number of supplementary analyses and reports have been elaborated and submitted for this procedure. Many workshop meetings have been held with the nominated experts. The hypothetical accidents have been analysed and a special report on these accidents has been submitted. The safety analyses report has been revised, taking into account the results and achievements reached during the ongoing licensing procedure. Parallel to these engineering activities outstanding in R and D work for the HTR-Module, e.g. in the field of fuel elements etc. has been continued. The HTR-Module has found worldwide interest. Respective activities are going on in Bangladesh, PR China, USSR, Indonesia etc. Relevant application studies have been carried out and/or initiated. (author). 15 refs, 16 figs

  18. HTR-10GT AMBs displacement sensor design

    International Nuclear Information System (INIS)

    Shi Zhengang; Zha Meisheng; Zhao Lei; Sun Zhuo

    2005-01-01

    The 10 MW high temperature gas-cooled test module reactor (HTR-10GT) with the core made of spherical fuel elements was designed and constructed by the Institute of Nuclear and New Energy Technology of Tsinghua University in China. In the HTR-10GT, turbo-compressor and generator rotors are connected by a flexible coupling. The rotors, restricted by actual instruments and working environment, must be supported without any contact and lubrication. Active magnetic bearing (AMB), known as its advantages over the conventional bearings., such as contact-free, no-lubricating and active damping vibration, is the best way to suspend and stabilize the position of rotors of HTR-10GT. Each rotor is suspended by two radial and one axial AMBs. The radial AMB's radial gap is 0.15 mm considering the gap of 0.4 mm between the compressor stator and blades in order to protect the compressor. The control system controls the rotor position to meet the required gaps between rotor and stator through windings current. All the position information concerning radial and axial AMB is generated by sensors for measuring the displacement of the levitated body. Some typical sensors, i.e. eddy current displacement sensor, capacitive displacement sensor, can provide position information, but, quite often, unsatisfactory anti-jamming, which is a key issue for AMB systems near generator and other electric devices in HTR-10GT. Therefore, a kind of new type sensor is designed to measure the radial and axial displacements and the vibration of the rotors. This paper focuses on the design characteristics of the HTR-10GT AMBs displacement sensors and introduction of the related experiments to demonstrate its performance. (authors)

  19. An HTR cogeneration system for industrial applications

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.; Heek, A.I. van; Kikstra, J.F.

    2001-01-01

    Because of its favourable characteristics of safety and simplicity the high-temperature reactor (HTR) could become a competitive heat source for a cogeneration unit. The Netherlands is a world leading country in the field of cogeneration. As nuclear energy remains an option for the medium and long term in this country, systems for nuclear cogeneration should be explored and developed. Hence, ECN Nuclear Research is developing a conceptual design of an HTR for Combined generation of Heat and Power (CHP) for the industry in and outside the Netherlands. The design of this small CHP-unit for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. The concept that was subject of this study, INCOGEN, used a 40 MW thermal pebble bed HTR and produced a maximum amount of electricity plus low temperature heat. The system has been improved to produce industrial quality heat, and has been renamed ACACIA. The output of this installation is 14 MW electricity and 17 tonnes of steam per hour, with a pressure of 10 bar and a temperature of 220 deg. C. The economic characteristics of this installation turned out to be much more favourable using modern data. The research work for this installation is embedded in a programme that has links to the major HTR projects in the world. Accordingly ECN participates in several IAEA Co-ordinated Research Programmes (CRPs). Besides this, ECN is involved in the South African PBMR-project. Finally, ECN participates in the European Concerted Action on Innovative HTR. (author)

  20. Neutronic feasibility design of a small long-life HTR

    International Nuclear Information System (INIS)

    Ding Ming; Kloosterman, Jan Leen

    2011-01-01

    Highlights: ► We propose the neutronic feasibility design of a small, long lifetime and transportable HTR. ► Comparison of cylindrical, annular and scatter cores of the small block-type HTR. ► The design of the scatter core effectively reduces the number of the fuel block and increases the lifetime and burnup of the reactor. - Abstract: Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U

  1. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  2. PCTR experiments with HTR lattice in MARIUS

    Energy Technology Data Exchange (ETDEWEB)

    Gambier, G; Estiot, J C; de Lapperent, D; Laponche, B; Luffin, J; Morier, F

    1972-06-15

    PCTR experiments have been carried out in Marius III with HTR tubular fuel, enriched to around 1% in order to reach K{sub infinity} = 1 and to reduce the mass of poison. Three poisons were used - Aluminium, Copper and Vanadium. The effect of air was measured and corrections were made to the results to allow the effect of delayed neutrons and the effect of axial heterogeneities. Interpretation was made with APOLLO. (auth)

  3. The HTR, applications, economics and environmental aspects

    International Nuclear Information System (INIS)

    Barnert, H.; Schad, M.; Candeli, H.

    1990-01-01

    The High Temperature Reactor (HTR), as the only nuclear system producing high temperature heat up to 1000 deg. C, offers a wide variety of applications. Besides electricity production, via steam turbines and in future via gas turbines, there is: District heat with high efficiency, long distance energy for urban energy supply, high pressure injection steam production for enhanced oil recovery, medium range temperature heat direct application in chemical and related industry and last not least, high temperature application for the refinement of fossil energy carriers. Recent results of studies and programmes will be presented: Near term applications are identified, e.g. refineries and alumina industry with smaller HTR units. Another large market is the production of hydrogen, methanol and ammonia on the basis of natural gas, the relevant technology has been developed up to the pilot scale. The refinement of fossil energy carriers, in particular of coal, is subject of the R+D programme in the cooperation between German industrial companies and the Nuclear Research Center. The results are very promising and will be explained in detail. This programme will be continued. Objectives are: improvement of the technology and of the economics as well as environmental aspects, e.g. the reduction of emissions of carbon-dioxid. The topics of the programme deal with the different apparatus, e.g. steam methane reformer, steam coal gasifier, intermediate heat exchanger and last not least, the process heat HTR. (author)

  4. An HTR cogeneration system for industrial application

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.; Van Heek, A.I.; Kikstra, J.F.

    1999-01-01

    Because of its favourable characteristics of safety and simplicity the high-temperature reactor (HTR) could become a competitive heat source for a cogeneration unit. The Netherlands is a world leading country in the field of cogeneration. As nuclear energy remains an option for the medium and long term in this country, systems for nuclear cogeneration should be explored and developed. Hence, ECN Nuclear Research is developing a conceptual design of an HTR for Combined generation of Heat and Power (CHP) for the industry in and outside the Netherlands. The design of this small CHP-unit for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. The concept that was subject of that study, INCOGEN, used a 40 MW thermal pebble bed HTR and produced a maximum amount of electricity plus low temperature heat. The system has been improved to produce industrial quality heat, and has been renamed ACACIA. The output of this installation is 14 MW electricity and 17 tonnes of steam per hour, with a pressure of 10 bar and a temperature of 220C. The economic characteristics of this installation turned out to be much more favourable using modern cost data. 15 refs

  5. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  6. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  7. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  8. The Energy Conversion Analysis of HTR Gas Turbine System

    International Nuclear Information System (INIS)

    Utaja

    2000-01-01

    The energy conversion analysis of HTR gas turbine system by hand calculation is tedious work and need much time. This difficulty comes from the repeated thermodynamic process calculation, both on compression or expansion of the cycle. To make the analysis faster and wider variable analyzed, HTR-1 programme is used. In this paper, the energy conversion analysis of HTR gas turbine system by HTR-1 will be described. The result is displayed as efficiency curve and block diagram with the input and output temperature of the component. This HTR-1 programme is developed by Basic language programming and be compiled by Visual Basic 5.0 . By this HTR-1 programme, the efficiency, specific power and effective compression of the amount of gas can be recognized fast. For example, for CO 2 gas between 40 o C and 700 o C, the compression on maximum efficiency is 4.6 and the energy specific is 18.9 kcal/kg, while the temperature changing on input and output of the component can be traced on monitor. This process take less than one second, while the manual calculation take more than one hour. It can be concluded, that the energy conversion analysis of the HTR gas turbine system by HTR-1 can be done faster and more variable analyzed. (author)

  9. Tardive dyskinesia and DRD3, HTR2A and HTR2C gene polymorphisms in Russian psychiatric inpatients from Siberia

    NARCIS (Netherlands)

    Al Hadithy, A. F. Y.; Ivanova, S. A.; Pechlivanoglou, P.; Semke, A.; Fedorenko, O.; Kornetova, E.; Ryadovaya, L.; Brouwers, J. R. B. J.; Wilffert, B.; Bruggeman, R.; Loonen, A. J. M.

    2009-01-01

    Background: Pharmacogenetics of tardive dyskinesia and dopamine D3 (DRD3), serotonin 2A (HTR2A), and 2C (HTR2C) receptors has been examined in various populations, but not in Russians. Purpose: To investigate the association between orofaciolingual (TDof) and limb-truncal dyskinesias (TDlt) and

  10. HTR-PM Safety requirement and Licensing experience

    International Nuclear Information System (INIS)

    Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

    2014-01-01

    HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

  11. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  12. Plutonium re-cycle in HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J. A.

    1974-03-15

    The study of plutonium cycles in HTRs using reprocessed plutonium from Magnox and AGR fuel cycles has shown that full core plutonium/uranium loadings are in general not feasible, burn-up is limited due the need for lower loadings of plutonium to meet reload core reactivity limits, on-line refueling is not practicable due to the need for higher burnable poison loadings, and low conversion rates in the plutonium-uranium cycles cannot be mitigated by axial loading schemes so that fissile make-up is needed if HTR plutonium recycle is desired.

  13. The HTR-10 project and its further development

    International Nuclear Information System (INIS)

    Xu Yuanhui

    2002-01-01

    The 10 MW High Temperature Gas-cooled Reactor-Test Module (termed as HTR-10) is one of key project in the National High Technology Research and Development Program (1986-2000). Main objectives for the HTR-10 are: (1). To acquire know-how to design, construct and operate the HTGRs, (2). To establish an experimental facility, (3). To demonstrate the inherent safety features of the Modular HTGR, (4). To test electricity and heat co-generation and closed cycle gas turbine technology and (5). To do research and development work for high temperature process heat application. The Institute of Nuclear Energy Technology (INET) of Tsinghua University was appointed as the leading institute to be responsible for design, license applications, construction and operation of the HTR-10. The HTR-10 technical design represents the features of HTR-Module design. After five years construction, installation and pre-operation the HTR-10 reached the criticality in December 2000. Up to now all of results on zero point experiments and fuel elements irradiation test are fine. China will continue to develop the high temperature gas-cooled reactor in the future using the HTR-10 base

  14. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Directory of Open Access Journals (Sweden)

    Przemyslaw Glaza

    Full Text Available Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3, showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ and an N-terminally truncated HtrA3S (ΔN-HtrA3S were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  15. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Science.gov (United States)

    Glaza, Przemyslaw; Osipiuk, Jerzy; Wenta, Tomasz; Zurawa-Janicka, Dorota; Jarzab, Miroslaw; Lesner, Adam; Banecki, Bogdan; Skorko-Glonek, Joanna; Joachimiak, Andrzej; Lipinska, Barbara

    2015-01-01

    Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3), showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ) and an N-terminally truncated HtrA3S (ΔN-HtrA3S) were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  16. Instrumentation of steam cycle HTR's up to 900 MWe

    International Nuclear Information System (INIS)

    Leithner, D.E.; Winkenbach, B.

    1982-06-01

    Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

  17. A 350 MW HTR with an annular pebble bed core

    International Nuclear Information System (INIS)

    Wang Dazhong; Jiang Zhiqiang; Gao Zuying; Xu Yuanhui

    1992-12-01

    A conceptual design of HTR-module with an annular pebble bed core was proposed. This design can increase the unit power capacity of HTR-Module from 200 MWt to 350 MWt while it can keep the inherent safety characteristics of modular reactor. The preliminary safety analysis results for 350 MW HTR are given. In order to solve the problem of uneven helium outlet temperature distribution a gas flow mixing structure at bottom of core was designed. The experiment results of a gas mixing simulation test rig show that the mixing function can satisfy the design requirements

  18. AREVA HTR concept for near-term deployment

    Energy Technology Data Exchange (ETDEWEB)

    Lommers, L.J., E-mail: lewis.lommers@areva.com [AREVA Inc., 2101 Horn Rapids Road, Richland, WA 99354 (United States); Shahrokhi, F. [AREVA Inc., Lynchburg, VA (United States); Mayer, J.A. [AREVA Inc., Marlborough, MA (United States); Southworth, F.H. [AREVA Inc., Lynchburg, VA (United States)

    2012-10-15

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  19. Two Phase Flow Stability in the HTR-10 Steam Generator

    Institute of Scientific and Technical Information of China (English)

    居怀明; 左开芬; 刘志勇; 徐元辉

    2001-01-01

    A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

  20. HTR plus modern turbine technology for higher efficiencies

    International Nuclear Information System (INIS)

    Barnert, H.; Kugeler, K.

    1996-01-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question 'can the HTR compete with high efficiencies?' is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab

  1. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  2. AREVA HTR concept for near-term deployment

    International Nuclear Information System (INIS)

    Lommers, L.J.; Shahrokhi, F.; Mayer, J.A.; Southworth, F.H.

    2012-01-01

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  3. HTR plus modern turbine technology for higher efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-08-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question `can the HTR compete with high efficiencies?` is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab.

  4. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  5. HTR-PM Progress and Further Commercial Deployment

    International Nuclear Information System (INIS)

    Wu, Frank

    2017-01-01

    Project Milestones: • 2004: industry investment agreement was signed • 2006: decided to use 2×250 MWt reactor modules with a 200 MWe steam turbine, became a key government R&D project • 2008: ATP was issued • 2012.12.9: FCD the first concrete poured. Chinese HTR development: HTR Roles in China - Power generation: supplement to LWR; repowering coal fired plants - Co-generation to supply steam - Hydrogen production

  6. Market potential of heat utilization of modular HTR in Japan

    International Nuclear Information System (INIS)

    Ide, Akira; Tasaka, Kanji.

    1993-01-01

    HTR is considered to be the most suitable reactor type to use in the field other than power generation. So it is useful to know market potential of this type of reactor in Japan to justify its development. This potential was estimated to be about 400 200MWt modular HTR reactors. This number will be double if the market of hydrogen is developed. (J.P.N.)

  7. HTR's role in process heat applications

    International Nuclear Information System (INIS)

    Kuhr, Reiner

    2008-01-01

    Advanced high-temperature nuclear reactors create a number of new opportunities for nuclear process heat applications. These opportunities are based on the high-temperature heat available, smaller reactor sizes, and enhanced safety features that allow siting close to process plants. Major sources of value include the displacement of premium fuels and the elimination of CO 2 emissions from combustion of conventional fuels and their use to produce hydrogen. High value applications include steam production and cogeneration, steam methane reforming, and water splitting. Market entry by advanced high-temperature reactor technology is challenged by the evolution of nuclear licensing requirements in countries targeted for early applications, by the development of a customer base not familiar with nuclear technology and related issues, by convergence of oil industry and nuclear industry risk management, by development of public and government policy support, by resolution of nuclear waste and proliferation concerns, and by the development of new business entities and business models to support commercialization. New HTR designs may see a larger opportunity in process heat niche applications than in power given competition from larger advanced light water reactors. Technology development is required in many areas to enable these new applications, including the commercialization of new heat exchangers capable of operating at high temperatures and pressures, convective process reactors and suitable catalysts, water splitting system and component designs, and other process-side requirements. Key forces that will shape these markets include future fuel availability and pricing, implementation and monetization of CO 2 emission limits, and the formation of international energy and environmental policy that will support initiatives to provide the nuclear licensing frameworks and risk distribution needed to support private investment. This paper was developed based on a plenary

  8. Introduction of HTR-PM Operation and Fuel Management System

    International Nuclear Information System (INIS)

    Liu Fucheng; Luo Yong; Gao Qiang

    2014-01-01

    There is a big difference between High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) and PWR in operation mode. HTR-PM is a continually refuelled reactor, and the operation and fuel management of it, which affect each other, are inseparable. Therefore, the analysis of HTR-PM fuel management needs to be carried out “in real time”. HTR-PM operation and fuel management system is developed for on-power refuelling mode of HTR-PM. The system, which calculates the core neutron flux and power distribution, taking high-temperature reactor physics analysis software-VSOP as a basic tool, can track and predict the core state online, and it has the ability to restructure core power distribution online, making use of ex-core detectors to correct and check tracking calculation. Based on the ability to track and predict, it can compute the core parameters to provide support for the operation of the reactor. It can also predict the operation parameters of the reactor to provide reference information for the fuel management.The contents of this paper include the development purposes, architecture, the main function modules, running process, and the idea of how to use the system to carry out HTR-PM fuel management. (author)

  9. The future of HTR development and market chances

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    In more than thirty years of development, the pebble bed high-temperature reactor has been brought to the threshold of commercial maturity. On the basis of the experience accumulated with the 15 MW AVR reactor and the THTR-300, unit sizes tailored to demand (HTR-500, modular HTR, GHR-10) will be developed for the electricity and heat markets of the future. The high-temperature reactor is a meaningful supplement to the proven line of light-water reactors and is particularly suitable for being exported to developing countries and industrial threshold countries because of its special technical and inherent safeguards properties. There is broad worldwide interest in the HTR, as is evidenced by several existing agreements on cooperation. It is for this reason that market chances are believed to exist for the HTR after the expected revival of the nuclear power market. ABB and Siemens therefore have decided to develop and market the HTR jointly in the future as a matter of long term strategy by working through a joint subsidiary, HTR-GmbH. (orig.) [de

  10. Why HTR/VHTR? A European point of view

    International Nuclear Information System (INIS)

    Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Ftitterer, M.; Van Heek, A.; Hittner, D.; Von Lensa, W.; Pirson, J.; Verrier, D.

    2008-01-01

    The (European) High Temperature Reactor Technology Network (HTR-TN) was created in 2000 by the main industrial and Research actors of nuclear energy in Europe for elaborating a strategy for developing advanced HTR technology towards industrial application and for taking initiatives for implementing this strategy, most particularly through the Euratom funded R and D programmes. HTR-TN members are convinced that the main market push for industrial deployment of a new generation of HTR will not come from utility needs for electricity generation, but from industrial process heat needs: even if HTR can be considered for satisfying particular niches of the electricity market, there will not be any incentive for utilities already experienced in the exploitation of large LWR to take the risk of a significant technology change, when no evident competitive edge would result from it. On the contrary, HTR is the sole nuclear system that can address heat needs of a large number of industrial processes that require a higher temperature than the temperature provided by all other types of industrial reactors. The possibility for HTR to address the industrial process heat market is a strong asset, as it opens to HTR a large market which is presently looking for solutions to reduce drastically CO 2 emissions, but at the same time it is a huge challenge: industrial exploitation of nuclear energy has been for the time being focused on electricity generation for which user requirements are relatively uniform. The versatility of process heat needs in terms of power, temperature, reliability, etc. will require a much larger flexibility of the nuclear heat source, which is not usual for nuclear industry, looking for competitiveness through standardisation. Therefore HTR-TN considers that the top priority innovation for HTR present development should not be missed: it is to demonstrate at an industrial scale the technical, industrial and economical feasibility of the coupling of a HTR with

  11. Means, methods and performances of the AREVA's HTR compact controls

    International Nuclear Information System (INIS)

    Banchet, J.; Guillermier, P.; Tisseur, D.; Vitali, M. P.

    2008-01-01

    In the AREVA's HTR development program, the reactor plant is composed of a prismatic core containing graphite cylindrical fuel elements, called compacts, where TRISO particles are dispersed. Starting from its past compacting process, the latter being revamped through the use of state of the art equipments, CERCA, 100% AREVA NP's subsidiary, was able to recover the quality of past compacts production. The recovered compacting process is composed of the following manufacturing steps: graphite matrix granulation, mix between the obtained granulates and particles, compacting and calcining at low pressure and temperature. To adapt this past process to new manufacturing equipments, non destructive examination tests were carried out to assess the compact quality, the latter being assessed via in house developed equipments and methods at each step of the design of experiments. As for the manufacturing process, past quality control methods were revamped to measure compact dimensional features (diameter, perpendicularity and cone effect), visual aspect, SiC layer failure fraction (via anodic disintegration and burn leach test) and homogeneity via 2D radiography coupled to ceramography. Although meeting quality requirements, 2D radiography method could not provide a quantified specification for compact homogeneity characterization. This limitation yielded the replacement of this past technique by a method based on X-Ray tomography. Development was conducted on this new technique to enable the definition of a criterion to quantify compact homogeneity, as well as to provide information about the distances in between particles. This study also included a comparison between simulated and real compacts to evaluate the accuracy of the technique as well as the influence of particle packing fraction on compact homogeneity. The developed quality control methods and equipments guided the choices of manufacturing parameters adjustments at the development stage and are now applied for

  12. HTR-TN achievements and prospects for future developments

    International Nuclear Information System (INIS)

    Hittner, D.; Angulo, C.; Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Futterer, M.A.; Van Heek, A.; Von Lensa, W.; Yvon, P.

    2011-01-01

    It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the 'Strategic Energy Technology Plan (SET-Plan)' recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to

  13. State of the art in HTR engineering and design

    International Nuclear Information System (INIS)

    Baust, E.

    1984-11-01

    The high-temperature reactor is an universally applicable energy source on the electricity and heat market, providing energy safely, compatible with the environment, and economically. The startup of the THTR-300, which will commence power generation in spring 1985, and the good results of the preparatory tests and studies for the subsequent plant, the HTR-500, created the required preconditions for the placing of an order to commence work to realize the first planning stage of the HTR-500. The order is expected to be placed within short. BBC/HRB has gained a reputation worldwide as the leading manufacturer of HTR plants. BBC/HRB has the know-how to offer HTR plants of various size over the entire capacity range between 100 and 600 MWe, or as twin-type plants up to 1200 MWe, their design being based on the THTR-300 reference plant. The HTR is an uncomplicated reactor system offering many advantages in terms of operation and safety. This reactor type therefore is the system of choice for energy generation for short-range energy supply. The system also is of interest as an export item, and hence is of significance to the economy and to employment policy. (orig.) [de

  14. Multicavity PCPVs for HTR and GCFR systems

    International Nuclear Information System (INIS)

    Eadie, D.Mc.D.

    1979-01-01

    There is little extra to report since the presentation of the paper 180/75 Multicavity PCPVs for HTR and GCFR Systems by P.L.T. Morgan and J.N. Bradbury at the International Conference on Experience in the Design, Construction and Operation of Prestressed Concrete Pressure Vessels and Containments for Nuclear Reactors at York, England, in September 1975. The paper presented at the York Conference demonstrated how a particular mode of behaviour could develop in a very local region between the pods and the external wall of a multicavity pressure vessel. Two main points emerge from the paper presented at York - 1. Local analysis for equilibrium of parts of the structure are as important as analysis of the general structural behaviour. With modern computer techniques, in which crack propagation and plasticity may be included, the development of local critical areas can be observed, but the idealisation of the structure has to be sufficiently refined and the cost will be high; 2. Criteria for acceptance of a design must be realistic and must be continually reviewed in the light of the trends of design philosophy. In conclusion, some pictures of model tests demonstrate the physical reality of the mode of failure described in the paper

  15. Microscopic thermal characterization of HTR particle layers

    International Nuclear Information System (INIS)

    Rochais, D.; Le Meur, G.; Basini, V.; Domingues, G.

    2008-01-01

    This paper presents thermal diffusivity measurements of HTR fuel particle pyrolytic carbon layers at room temperature. The photoreflectance microscopy (PM) technique is used to characterize particle layers at a microscopic scale. Nevertheless, buffer layer needs a particular analysis due to its porous structure. Indeed, measurements by PM on this material only permit to obtain the thermal diffusivity of the solid skeleton, whose homogeneous zones surface does not exceed 100 μm 2 . These characteristics make, on the one hand, delicate the use of PM, and on the other hand, require the use of a numerical homogenization technique. This model takes into account the properties of gas confined in the pores, to simulate the conduction heat flux traveling through the layer in relation with its microstructure and to estimate an effective thermal conductivity of the entire layer. This approach is validated by infrared microscopy measurement of the effective thermal diffusivity of the especially elaborated thicker buffer layer. Last, the first tests to characterize the silicon carbide layer are presented

  16. Results and future programme of HTR's study

    International Nuclear Information System (INIS)

    Mursid Djokolelono; Soedyartomo Soentono

    1990-01-01

    Study on the application of HTRs for the enhanced oil recovery in the Duri oil field (Sumatra, Indonesia) was performed in 1986/1987. The economic and technological advantages over crude burning option were identified. Crude oil prices, HTR capital costs, discount rates and company's income structure represented dominant parameters. Further sensitivity calculations on important economic parameters were obtained to reflect the condition of 1988. This nuclear option was also incorporated in the energy planning study for the whole of Indonesia using the MARKAL model, and resulted in the conditions of its applicability. The scenarios chosen in this MARKAL study were high and low GDP growth rate, whereas the criteria chosen were the minimum cost with and without a predetermined policy of reduced domestic use of oil. In the high scenario the HTRs as well as the natural gas options could not compete against the low cost boilers with crude-oil fuel. But in the case of reduced domestic oil use the HTRs came out to supplement the crudeburning boilers starting in the sixth five year plan (1994-999), even earlier than the natural gas option. The authors further discuss the industrial environment, in relation to the regional development, the possible local participation, as well as the plan to materialize the merits of this novel application. (author)

  17. Results and future programme of HTR's study

    Energy Technology Data Exchange (ETDEWEB)

    Djokolelono, Mursid; Soentono, Soedyartomo [National Atomic Energy Agency (Indonesia)

    1990-07-01

    Study on the application of HTRs for the enhanced oil recovery in the Duri oil field (Sumatra, Indonesia) was performed in 1986/1987. The economic and technological advantages over crude burning option were identified. Crude oil prices, HTR capital costs, discount rates and company's income structure represented dominant parameters. Further sensitivity calculations on important economic parameters were obtained to reflect the condition of 1988. This nuclear option was also incorporated in the energy planning study for the whole of Indonesia using the MARKAL model, and resulted in the conditions of its applicability. The scenarios chosen in this MARKAL study were high and low GDP growth rate, whereas the criteria chosen were the minimum cost with and without a predetermined policy of reduced domestic use of oil. In the high scenario the HTRs as well as the natural gas options could not compete against the low cost boilers with crude-oil fuel. But in the case of reduced domestic oil use the HTRs came out to supplement the crudeburning boilers starting in the sixth five year plan (1994-999), even earlier than the natural gas option. The authors further discuss the industrial environment, in relation to the regional development, the possible local participation, as well as the plan to materialize the merits of this novel application. (author)

  18. Status of the HTR programme in France

    International Nuclear Information System (INIS)

    Ballot, B.; Gauthier, J.C.; Hittner, D.; Lebrun, J.Ph.; Lecomte, M.; Carre, F.; Delbecq, J.M.

    2007-01-01

    AREVA is convinced that HTR (High Temperature Reactor) is not in competition with large LWRs for electricity generation, and that its actual added value is its potential for addressing cogeneration and industrial process heat production. Therefore AREVA launched in 2004 the ANTARES programme for a pre-conceptual design study, with the support of EDF and together with a large research and development programme needed for the design in close collaboration with Cea. The pre-conceptual phase was finalized end of 2006. The specific feature of AREVA's concept, which distinguishes it from other ones, is its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important being the design flexibility to adapt readily to combined heat and power applications, with a standardised nuclear heat source as independent as possible of the versatile process heat applications with which it is coupled. Standardisation should expedite licensing. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source free of greenhouse gases emissions. The reactor module produces 600 MWth which can be split into the required process heat, the remaining power driving an adapted prorated electric plant. Depending on the process heat temperature and power needs, up to 80 % of the nuclear heat is converted into useful energy

  19. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

    Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

  20. Overview of Japanese seismic research program for HTR

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1978-07-01

    In order to obtain the license for construction and operation of HTR developed and introduced into Japan, it is necessary to assure integrity of reactor structures and the capability of reactor shutdown and maintain safety shutdown for the seismic design condition. Because Japanese land is located in relatively high seismacity zone, when an excessive earthquake would occur, the public and plant personnel should be protected from radiation hazard. For the above reason, many efforts of seismic research and development for HTR have been made at institutes and companies in Japan. In the paper, descriptions are: (1) Present status of development and construction plans of HTR, (2) guideline of aseismic design, (3) need of aseismic research, (4) present status of research and development, (5) future plan. (auth.)

  1. Thermodynamic correlations for the accident analysis of HTR's

    International Nuclear Information System (INIS)

    Rehm, W.; Jahn, W.; Finken, R.

    1976-12-01

    The thermal properties of Helium and for the case of a depressurized primary circuit, various mixtures of primary cooling gas were taken into consideration. The temperature dependence of the correlations for the thermal properties of the graphite components in the core and for the structural materials in the primary circuit are extrapolated about normal operation conditions. Furthermore the correlations for the effective thermal conductivity, the heat transfer and pressure drop are described for pebble bed HTR's. In addition some important heat transfer data of the steam generator are included. With these correlations, for example accident sequences with failure of the afterheat removal systems are discussed for pebble bed HTR's. It is concluded that the transient temperature behaviour demonstrates the inherent safety features of the HTR in extreme accidents. (orig.) [de

  2. HTR-E project. High-temperature components and systems

    International Nuclear Information System (INIS)

    Breuil, E.; Exner, R.

    2002-01-01

    The HTR-E European project (four years project) is proposed for the 5th Framework Programme and concerns the technical developments needed for the innovative components of a modern HTR with a direct cycle. These components have been selected with reference to the present projects (GT-MHR, PBMR): (1) the helium turbine, the recuperator heat exchanger, the electro-magnetic bearings and the helium rotating seal; (2) the tribology. Sliding innovative components in helium environment are particularly concerned. (3) the helium purification system. Recommendations on impurities contents have to be provided in accordance with the materials proposed for the innovative components. The main outcomes expected from the HTR-E project are the design recommendations and identification of further R and D needs for these components. This will be based: (1) on experience feedback from European past helium test loops and reactors; (2) on design studies, thermal-hydraulic and structural analyses; (3) and on experimental tests

  3. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  4. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  5. Risk assessment of small-sized HTR with pebble-bed core

    International Nuclear Information System (INIS)

    Kroeger, W.; Mertens, J.; Wolters, J.

    1987-01-01

    Two recent concepts of small-sized HTR's (HTR-Modul and HTR-100) were analysed regarding their safety concepts and risk protection. In neither case do core cooling accidents contribute to the risk because of the low induced core temperatures. Water ingress accidents dominate the risk in both cases by detaching deposited fission products which can be released into the environment. For these accident sequences no early fatalities and practically no lethal case of cancer were computed. Both HTR concepts include adequate precautionary measures and an infinitely small risk according to the usual standards. The safety concepts make express use of the specific inherent safety features of pebble-bed HTR's. (orig.)

  6. HTR process heat applications, status of technology and economical potential

    International Nuclear Information System (INIS)

    Barnet, H.

    1997-01-01

    The technical and industrial feasibility of the production of high temperature heat from nuclear fuel is presented. The technical feasibility of high temperature heat consuming processes is reviewed and assessed. The conclusion is drawn that the next technological step for pilot plant scale demonstration is the nuclear heated steam reforming process. The economical potential of HTR process heat applications is reviewed: It is directly coupled to the economical competitiveness of HTR electricity production. Recently made statements and pre-conditions on the economic competitiveness in comparison to world market coal are reported. (author). 8 figs

  7. Relevant safety issues in designing the HTR-10 reactor

    International Nuclear Information System (INIS)

    Sun Yuliang; Xu Yuanghui

    2001-01-01

    The HTR-10 is a 10 MWth pebble bed high temperature gas cooled reactor being constructed as a research facility at the Institute of Nuclear Energy Technology. This paper discusses design issues of the HTR-10 which are related to safety. It addresses the safety criteria used in the development and assessment of the design, the safety important systems, and the safety classification of components. It also summarises the results of safety analysis, including the approach used for the radioactive source term, as well as the approach to containment design. (author)

  8. Energy analysis of control rod drive mechanism in HTR-10

    International Nuclear Information System (INIS)

    Bo Hanliang; Wu Yuanqiang

    2000-01-01

    This paper presents a theoretical model for the control rod drive mechanism for the 10 MW High Temperature Gas Cooled Reactor (HTR-10) and analyzes accidents which may occur in the drive mechanism, for example, chain break, coupling damage and other damage scenarios. The results show that the matching problem between buffer capability and coupling strength is the main reason for coupling damage; increased temperatures would reduce eddy damping and cause a mismatch between buffer capability and coupling strength; and the displacement of the buffer spring will affect the coupling force. The results provide a theoretical basis for the design of the control rod drive mechanism for HTR-10

  9. Engineering and licensing progress of the HTR-Module

    Energy Technology Data Exchange (ETDEWEB)

    Weisbrodt, I A

    1988-07-01

    This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress.

  10. Engineering and licensing progress of the HTR-Module

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1988-01-01

    This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress

  11. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

  12. Overview of Japanese seismic research program for HTR

    International Nuclear Information System (INIS)

    Ikushima, T.

    1978-01-01

    In order to obtain the license for construction and operation of HTR developed in and/or introduced into Japan, it is necessary to insure the integrity of reactor structures and the capability of reactor shutdown and the maintenance of safety shutdown for the seismic design condition. Because Japan is located in relatively high seismicity zone, even when an excessive earthquake would occur, the public and plant personnel should be protected from radiation hazard. The report describes the following: (1) present status of development and construction plan of HTR, (2) guideline of aseismic design, (3) need of aseismic research, (4) present status of research and development, and (5) future plans

  13. Transient behaviour of small HTR for cogeneration

    International Nuclear Information System (INIS)

    Verkerk, E.C.; Van Heek, A.I.

    2000-01-01

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MWe electricity combined with 17 t/h of high temperature steam (220 deg C, 10 bar) with a pebble-bed high temperature reactor directly coupled with a helium compressor and a helium turbine. The design of this small CHP unit that is used for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. Thermal hydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package PANTHER/DIREKT and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This coupling offers a more realistic simulation of the entire system, since it removes the necessity of forcing boundary conditions on the simulation models at the data transfer points. In this paper, the models used for the dynamic components of the energy conversion system are described, and the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system are shown. Several transient cases that are representative as operational transients for an HTR will be discussed, including one representing a load rejection case that shows the functioning of the control system, in particular the bypass valve. Another transient is a load following

  14. A new impetus for developing industrial process heat applications of HTR in europe - HTR2008-58259

    International Nuclear Information System (INIS)

    Hittner, D.; De Groot, S.; Griffay, G.; Yvon, P.; Pienkowski, L.; Ruer, J.; Angulo, C.; Laquaniello, G.

    2008-01-01

    Due to its high operating temperature (up to 850 deg. C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R and D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible

  15. ARCHER HTR Technology in support of a Coal to Liquid Process – An Economic Feasibility View

    International Nuclear Information System (INIS)

    Stoker, P.W.; Fick, J.I.J.; Conradie, F.H.

    2014-01-01

    The paper considers the economics of coupling a European developed HTR (as conceptualized by project ARCHER) to a Coal-to-Liquid (CTL) process as typically used by Sasol, the biggest Coal-to-Liquid (CTL) producer in the world. The approach followed was to create a techno-economic baseline for an existing CTL process using mass and energy balances determined with Aspen Plus chemical modelling software. The economic performance of a typical 80,000 barrels per day synthetic crude oil plant was determined from first principles. The techno-economic baseline model was validated with reference to published product output data and audited financial results of a Sasol CTL plant located at Secunda, South Africa, as reported for the 2011 financial year. A number of schemes were identified to couple the European HTR plant to the CTL case study. Two schemes were studied in detail, while the remaining coupling schemes will be studied as part of the follow-on project NC2I-R (Nuclear Cogeneration Industrial Initiative – Research). Two Key Performance Indices were of interest, namely the Internal Rate of Return of a Nuclear supported CTL plant and the reduction of CO_2 emissions. The case where nuclear co-generation replaced electrical power bought from the grid, and also replaced all the steam currently produced by the burning coal with nuclear steam, yielded interesting conclusions: • The case study plant would need a total of 16 HTRs, each with a capacity of 265 MWth. • The coupling scheme would reduce CO_2 emissions by approximately 14.5 million ton/annum or 51 % of the current emissions of a 80,000 bbl/d plant. • The economic feasibility challenge for large scale deployment of nuclear energy in a Coal-to-Liquid application - where steam and electricity are to be generated from Nuclear energy, is to construct such a facility at an all -inclusive overnight cost not exceeding $3400/kWe. (author)

  16. Would HTR be suitable for application in the Netherlands?

    International Nuclear Information System (INIS)

    Heek, A.I. van.

    1994-08-01

    The modular HTR may be a reactor type, which would have sufficient societal support to be constructed in the Netherlands. The economic approach would be fundamentally different from that applied in present nuclear technology. In a national research program this is being investigated. (orig.)

  17. Approach to equilibrium calculations for the dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-06-10

    The calculational methods and the model used in representing the core and the fuel management operations are described. Different layouts of the first core and approach to equilibrium schemes for the Dragon HTR design are investigated. A simple fuelling modus is found and the tchnological and economical implications are discussed in detail.

  18. On Power Refueling Management of HTR-PM

    International Nuclear Information System (INIS)

    Sun Furui; Luo Yong; Gao Qiang

    2014-01-01

    The refueling management is an important work of nuclear power plant , directly affecting its safety and economy. At present, the ordinary commercial pressurized water reactor (PWR) nuclear power plant has developed more mature in the refueling management, and formed a set of relatively complete system and methods.The High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) has significant differences with the ordinary PWR nuclear power plant in the fuel morphology and the refueling mode. It adopts the spherical fuel element and the on-power refueling. Therefore, the HTR-PM refueling management has its own unique characteristics, but currently there is no mature experience to use for reference across the world. This paper gives a brief introduction to the HTR-PM on power refueling management, including the refueling management system construction, the refueling strategy, the fuel element internal transportation,charging and discharging, etc. It aims at finding the befitting HTR-PM refueling management methods in view of its own unique characteristics in order to ensure the orderly development of the refueling management and the refueling safety. (author)

  19. Profiles of facilities used for HTR research and testing

    International Nuclear Information System (INIS)

    1980-05-01

    This report contains a current description of facilities supporting HTR research and development submitted by countries participating in the IWGFR. It has the purpose of providing an overview of the facilities available for use and of the types of experiments that can be conducted therein

  20. Waste heat of HTR power stations for district heating

    International Nuclear Information System (INIS)

    Bonnenberg, H.; Schlenker, H.V.

    1975-01-01

    The market situation, the applied techniques, and the transport, for district heating in combination with HTR plants are considered. Analysis of the heat market indicates a high demand for heat at temperatures between 100 and 150 0 C in household and industry. This market for district heating can be supplied by heat generated in HTR plants using two methods: (1) the combined heat and power generation in steam cycle plants by extracting steam from the turbine, and (2) the use of waste heat of a closed gas turbine cycle. The heat generation costs of (2) are negligible. The cost for transportation of heat over the average distance between existing plant sites and consumer regions (25 km) are between 10 and 20% of the total heat price, considering the high heat output of nuclear power stations. Comparing the price of heat gained by use of waste heat in HTR plants with that of conventional methods, considerable advantages are indicated for the combined heat and power generation in HTR plants. (author)

  1. Solution of multiple circuits of steam cycle HTR system

    International Nuclear Information System (INIS)

    Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

    2014-01-01

    In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

  2. The Dragon project and high temperature reactor (HTR position)

    International Nuclear Information System (INIS)

    Shepherd, L.

    1981-01-01

    After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

  3. HTR-2002: Proceedings of the conference on high temperature reactors

    International Nuclear Information System (INIS)

    2002-01-01

    High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

  4. Experiments in MARIUS on HTR tubular fuel with loose particles

    Energy Technology Data Exchange (ETDEWEB)

    Bosser, R; Langlet, G

    1972-06-15

    The work described on HTR tubular fuel with loose particles is the first part of a program in three points. The cell is the same in the three experiments, only particles in the fuel container are changed. The aim of the experiment is to achieve the buckling in a critical facility. A description of the techniques of measurements, calculations, and results are presented.

  5. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  6. The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

    2016-09-15

    A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

  7. Maternal HtrA3 optimizes placental development to influence offspring birth weight and subsequent white fat gain in adulthood.

    Science.gov (United States)

    Li, Ying; Salamonsen, Lois A; Hyett, Jonathan; Costa, Fabricio da Silva; Nie, Guiying

    2017-07-04

    High temperature requirement factor A3 (HtrA3), a member of the HtrA protease family, is highly expressed in the developing placenta, including the maternal decidual cells in both mice and humans. In this study we deleted the HtrA3 gene in the mouse and crossed females carrying zero, one, or two HtrA3-expressing alleles with HtrA3 +/- males to investigate the role of maternal vs fetal HtrA3 in placentation. Although HtrA3 -/- mice were phenotypically normal and fertile, HtrA3 deletion in the mother resulted in intra-uterine growth restriction (IUGR). Disorganization of labyrinthine fetal capillaries was the major placental defect when HtrA3 was absent. The IUGR caused by maternal HtrA3 deletion, albeit being mild, significantly altered offspring growth trajectory long after birth. By 8 months of age, mice born to HtrA3-deficient mothers, independent of their own genotype, were significantly heavier and contained a larger mass of white fat. We further demonstrated that in women serum levels of HtrA3 during early pregnancy were significantly lower in IUGR pregnancies, establishing an association between lower HtrA3 levels and placental insufficiency in the human. This study thus revealed the importance of maternal HtrA3 in optimizing placental development and its long-term impact on the offspring well beyond in utero growth.

  8. State of the Art of helium heat exchanger development for future HTR-projects - HTR2008-58146

    International Nuclear Information System (INIS)

    Esch, M.; Juergens, B.; Hurtado, A.; Knoche, D.; Tietsch, W.

    2008-01-01

    In Germany two HTR nuclear power plants had been built and operated, the AVR-15 and the THTR-300. Also various projects for different purposes in a large power range had been developed, The AVR-15, an experimental reactor with a power output of 15 MWel was operated for more than 20 years with excellent results. The THTR-300 was designed as a prototype demonstration plant with 300 MWel and should be the technological basis for the entire future reactor line. The THTR-300 was prematurely shut down and decommissioned because of political reasons. But because of the accompanying comprehensive R and D program and the operation time of about 5 years, the technology was proved and essential operational results were gained. The AVR steam generator was installed above the reactor core. The six THTR heat exchangers were arranged circularly around the reactor core, Both heat exchanger systems have been operated successfully and furthermore acted as a residual heat removal system. The technology knowledge and experience gained on these existing HTR plants is still available at Westinghouse Electric Germany GmbH since Westinghouse is one of the legal successors of the former German HTR companies. As a follow-up project of THTR, the HTR-500 was developed and designed up to the manufacturing stage. For this plant additionally to the 8 steam generators, two residual heat removal heat exchangers were foreseen. These were to be installed in a ring around the reactor core. All these HTRs were designed for the generation of electricity using a steam cycle. Extensive research work has also been done for advanced applications of HTR technology e.g. using a direct cycle within the HHT project or generating process heat within the framework of the PNP project, Because of the critical attitude of the German government to the nuclear power in the past 20 years in Germany there was only a very limited interest in the further development of the HTR technology. As a consequence of the German

  9. Financing models for HTR plants: Co-financing, counter trade, joint ventures

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    Structure and volume of investment cost for HTR nuclear power plants are different in comparison to other types of nuclear power plants. Even if the share of local participation is in comparable order of magnitude to other nuclear power plants, the required technical infrastructure for HTR plants is more suitable for existing and still practised technologies in countries which are in development processes. These HTR specific features offer special possibilities in HTR project financing. Various models are discussed in respect of the special HTR situation. Even if it is not possible to point out in a general manner the best solution - due to national, local and time dependant situations - this paper discusses the HTR specific impacts to buyer's credit financing, supplier's credit financing, barter trades or joint ventures and combined financing. (author). 4 refs, 9 figs

  10. Ankle replacement

    Science.gov (United States)

    Ankle arthroplasty - total; Total ankle arthroplasty; Endoprosthetic ankle replacement; Ankle surgery ... Ankle replacement surgery is most often done while you are under general anesthesia. This means you will ...

  11. IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: High Temperature Reactor Studies, including experiments in critical facilities or in prototypes have been carried out in the past. Information gathered, experience gained and experimental data produced are of value for the development of future advanced HTRs. For the purpose of knowledge, competence, information preservation and management, computer readable archives have been established. The present archive includes several relevant documents relative to the following: - Graphite Moderated Critical Facility, CESAR at Cadarache. Dragon Countries Physics Meetings (DCPM); - OTTO Pebble Bed Reactors; - Gulf - HTGR Experiments; - Zero Power MARIUS Reactor; - Pebble-bed KAHTER Critical Facility; - Helium Cooled Fast Reactor Assessment Studies; - Gas Cooled Reactor Technology Safety and Siting; - Initial Evaluation of the Gas-Turbine Modules HTGCR; - A report on Nuclear Graphite; - AVR Reactor Juelich (new in version 02); - HTR IAEA proceedings (new in version 02); - Studies at IRI Delft(new in version 02); - Studies and experiments at PSI Villigen (new in version 02); 2 - Related or auxiliary information: IRPHE-DRAGON-DPR, high Temperature Reactor Dragon Project, Primary Documents NEA-1726/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  12. The chlamydial periplasmic stress response serine protease cHtrA is secreted into host cell cytosol

    Directory of Open Access Journals (Sweden)

    Flores Rhonda

    2011-04-01

    Full Text Available Abstract Background The periplasmic High Temperature Requirement protein A (HtrA plays important roles in bacterial protein folding and stress responses. However, the role of chlamydial HtrA (cHtrA in chlamydial pathogenesis is not clear. Results The cHtrA was detected both inside and outside the chlamydial inclusions. The detection was specific since both polyclonal and monoclonal anti-cHtrA antibodies revealed similar intracellular labeling patterns that were only removed by absorption with cHtrA but not control fusion proteins. In a Western blot assay, the anti-cHtrA antibodies detected the endogenous cHtrA in Chlamydia-infected cells without cross-reacting with any other chlamydial or host cell antigens. Fractionation of the infected cells revealed cHtrA in the host cell cytosol fraction. The periplasmic cHtrA protein appeared to be actively secreted into host cell cytosol since no other chlamydial periplasmic proteins were detected in the host cell cytoplasm. Most chlamydial species secreted cHtrA into host cell cytosol and the secretion was not inhibitable by a type III secretion inhibitor. Conclusion Since it is hypothesized that chlamydial organisms possess a proteolysis strategy to manipulate host cell signaling pathways, secretion of the serine protease cHtrA into host cell cytosol suggests that the periplasmic cHtrA may also play an important role in chlamydial interactions with host cells.

  13. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  14. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  15. Calculation of HTR-10 first criticality with MVP

    International Nuclear Information System (INIS)

    Xie Jiachun; Yao Lianying

    2015-01-01

    The first criticality of 10 MW pebble-bed high temperature gas-cooled reactor-test module (HTR-10) was calculated with MVP. According to the characteristics of HTR-10, the Statistical Geometry Model of MVP was employed to describe the random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core. Compared with previous results from VSOP and MCNP, the MVP results with JENDL-3.3 library were little more different, but the results with ENDF/B-Ⅵ.8 library were very close. The relative errors were less than 0.7%, compared with the first criticality experimental results. The study shows that MVP could be used in the physics calculations for pebble bed high temperature gas-cooled reactors. (authors)

  16. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  17. Turbo-machine deployment of HTR-10 GT

    International Nuclear Information System (INIS)

    Zhu Shutang; Wang Jie; Zhang Zhengming; Yu Suyuan

    2005-01-01

    As a testing project of gas turbine modular High Temperature Gas-cooled Reactor (HTGR), HTR-10GT has been studied and developed by Institute of Nuclear and New Energy Technology (INET) of Tsinghua University after the success of HTR-10 with steam turbine cycle. The main purposes of this project are to demonstrate the gas turbine modular HTGR, to optimize the deployment of Power Conversion Unit (PCU) and to verify the techniques of turbo-machine, operating modes and controlling measures. HTR-10GT is concentrated on the PCU design and the turbo-machine deployment. Possible turbo-machine deployments have been investigated and two of them are introduced in this paper. The preliminary design for the turbo-machine of HTR-10GT is single-shaft of vertical layout, arranged by the side of the reactor and the turbo-compressor rotary speed was selected to be 250 s -1 (15000 r/min) by considering the efficiency of turbo-compressor blade systems, the strength conditions and the mass and size characteristics of the turbo-compressor. The rotor system will be supported by electromagnetic bearings (EMBs) to curb the possible pollutions of the primary loop. Of all the components in this design, the high speed turbo-generator seems to be a world-wide technical nut. As an alternative design, a gearbox complex is used to reduce the rotary speed from the turbo-compressor 250 s -1 to 50 s -1 so that the ordinary generator can be used. (authors)

  18. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  19. Intermediate heat exchanger for HTR process heat application

    International Nuclear Information System (INIS)

    Crambes, M.

    1980-01-01

    In the French study on the nuclear gasification of coal, the following options were recommended: Coal hydrogenation, the hydrogen being derived from CH 4 reforming under the effects of HTR heat; the use of an intermediate helium circuit between the nuclear plant and the reforming plant. The purpose of the present paper is to describe the heat exchanger designed to transfer heat from the primary to the intermediate circuit

  20. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  1. Strengths, weaknesses, opportunities and threats for HTR deployment in Europe

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Kugeler, Kurt; Fütterer, Michael A.

    2014-01-01

    High temperature nuclear reactors are a technology, of which early versions were demonstrated in the 1960s–1980s in Germany (AVR, THTR) and the United States (Peach Bottom, Fort St. Vrain). HTRs were initially designed for high temperature, high efficiency electricity generation but the technology, the market and the targeted applications have evolved since then to address industrial cogeneration and new operational conditions (in particular new safety regulations). This paper intends to analyse the latest status of HTR today, as regards their intrinsic strengths and weaknesses and their external context, whether positive (opportunities) or negative (threats). Different dimensions are covered by the analysis: technology status, results from R and D programmes (especially in Europe), competences and skills, licensing aspects, experience feedback from demonstrator operation (in particular in Germany), economic conditions and other non-technical aspects. Europe has a comprehensive experience in the field of HTR with capabilities in both pebble bed and prismatic designs (R and D, engineering, manufacturing, operation, dismantling, and the full fuel cycle). Europe is also a promising market for HTR as the process heat market is large with good industrial and cogeneration infrastructures. The analysis of the European situation is to a good deal indicative for the global potential of this technology

  2. Objectives for an HTR R and D physics programme

    Energy Technology Data Exchange (ETDEWEB)

    Johnstone, I; Scott, J A

    1973-10-15

    The paper reviews important objectives for an HTR R and D programme and the importance of particular characteristics for safety and reactor performance is discussed. Uncertainties in the physics characteristics influence reactor design through the inclusion of design margins and contingency allowances and may even eliminate certain design variants. The paper discusses quantitatively the relationship between some important uncertainties and reactor design and operating costs and derives targets for the precision with which it should be possible to compute the corresponding physics characteristics. To extrapolate to power reactor conditions, the need for a comprehensive computational scheme validated by an adequate experimental programme is emphasised. The reduction in uncertainty as the theoretical and experimental reactor physics development proceeds in order to meet the desired target uncertainty is illustrated with respect to the UK's programme in support of the low enriched HTR concept. The current situation for this concept is discussed and compared briefly with that for the Th cycle HTR for which somewhat less zero energy experimental data are available. (auth)

  3. Survey of appropriate endothermic processes for association with the HTR

    International Nuclear Information System (INIS)

    Brown, G.; Harrison, G.E.; Gent, C.W.; Plummer, J.

    1975-01-01

    Emphasis is placed on association of the HTR system as a heat source with chemical processes requiring temperatures up to 850 to 900 0 C, corresponding to a reactor coolant temperature of 950 0 C, though processes requiring temperatures up to 1100 0 C and above are reviewed. Particular attention is given to processes for the production of hydrogen-containing gases, including coal/lignite gasification which has been the subject of a recent study. Rising fuel prices make the HTR an attractive proposition if design concepts and materials can be developed to match the requirements. Other appropriate endothermic processes considered are oil processing, including tar sands and shales, and also energy production. Since the full temperature range of the reactor system must be utilised mention is made of low grade heat uses. Even very large chemical works have relatively small energy requirement by nuclear heat standards and adoption of the HTR as a heat source is likely only if it is associated with a large chemical/metallurgical complex or with the processing of a natural resource. (author)

  4. Analysis on First Criticality Benchmark Calculation of HTR-10 Core

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

    HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

  5. 7th International Topical Meeting on High Temperature Reactor Technology: The modular HTR is advancing towards reality. Papers and Presentations

    International Nuclear Information System (INIS)

    2014-01-01

    HTR2014 aimed at providing an international platform for researchers, engineers and industrial professionals to share their innovative ideas, valuable experience and recent progresses on high temperature gas-cooled reactor (HTR) and its application technologies.

  6. EC-funded project (HTR-L) for the definition of a European safety approach for HTR's

    International Nuclear Information System (INIS)

    Ehster, S.; Dominguez, M.T.; Coe, I.; Brinkmann, G.; Lensa, W. von; Mheen, W. van der; Alessandroni, C.; Pirson, J.

    2002-01-01

    The inherent safety features of the HTRs make events leading to severe core damage highly unlikely and constitute the main differentiating aspects compared to LWRs. While a known and stable regulatory environment has long been established for Light Water Reactors, a different approach is necessary for the licensing of HTR based power plants. Among the R and D projects funded by the European Commission for HTR reactors, the HTR-L project is dedicated to the definition of a common and coherent European safety approach and the identification of the main licensing issues for the licensing framework of the Modular HTRs. Other specific objectives of this project are : To develop a methodology to classify the accidental conditions; To define the preliminary requirements for the confinement of radioactive products and to assess the need for a 'conventional' containment structure; To establish a SSC (2) classification and to define the rules for equipment qualification; To identify the key issues that need to be addressed in the licensing process of the HTRs; To organize a workshop with the concerned Safety Authorities at the end of the project. This paper will explain the project objectives and its final expected outcomes. (author)

  7. Enhancing the Sustainability of Nuclear Power: the Pebble Bed HTR in Deep Burn Mode

    International Nuclear Information System (INIS)

    Da Cruz, D.F.; De Haas, J.B.M.; Van Heek, A.I.

    2004-01-01

    The scenario of a utility in an industrialized country starting new nuclear construction with a single PBMR reactor has been considered. To make the new construction project acceptable by government and society, a maximum effort to obtain sustainability (i.e. minimization of resource use and waste production) will have to be shown. Therefore the usual open cycle for HTR has been abandoned, and the spent fuel will be reprocessed once. The long-lived transuranic (TRU) elements Pu, Np, Am and Cm are all re-fabricated into so-called transmutation fuel elements, and loaded back into the same reactor, in our case a 110 MWe PBMR with low-enriched uranium cycle. In this study, the reactor physical prospects have been investigated: to what extent the amount of TRU could be reduced. In this way, 75% of the initial amount of TRU waste is being destructed, while the time span in which the waste is more radio-toxic than uranium ore is being reduced to one-third. Also, the amount of fresh driver fuel needed is decreases by 25%. A preliminary cost analysis has been performed as well. It shows that there is also a cost advantage of operating the reactor in Deep Burn mode in industrialized countries, where the waste storage fees charged per volume are relatively high. (authors)

  8. In-situ hybridization based quantification of hTR: a possible biomarker in malignant melanoma

    DEFF Research Database (Denmark)

    Vagner, Josephine; Steiniche, Torben; Stougaard, Magnus

    2015-01-01

    thickness suggesting that hTR might be a valuable biomarker in MM. Furthermore, as ISH-based detection requires presence of both hTR and the reverse transcriptase (hTERT) it might be an indicator of active telomerase and thus have future relevance as a predictive biomarker for anti-telomerase treatment....

  9. Evolution of mitochondrial cell death pathway: Proapoptotic role of HtrA2/Omi in Drosophila

    International Nuclear Information System (INIS)

    Igaki, Tatsushi; Suzuki, Yasuyuki; Tokushige, Naoko; Aonuma, Hiroka; Takahashi, Ryosuke; Miura, Masayuki

    2007-01-01

    Despite the essential role of mitochondria in a variety of mammalian cell death processes, the involvement of mitochondrial pathway in Drosophila cell death has remained unclear. To address this, we cloned and characterized DmHtrA2, a Drosophila homolog of a mitochondrial serine protease HtrA2/Omi. We show that DmHtrA2 normally resides in mitochondria and is up-regulated by UV-irradiation. Upon receipt of apoptotic stimuli, DmHtrA2 is translocated to extramitochondrial compartment; however, unlike its mammalian counterpart, the extramitochondrial DmHtrA2 does not diffuse throughout the cytosol but stays near the mitochondria. RNAi-mediated knock-down of DmHtrA2 in larvae or adult flies results in a resistance to stress stimuli. DmHtrA2 specifically cleaves Drosophila inhibitor-of-apoptosis protein 1 (DIAP1), a cellular caspase inhibitor, and induces cell death both in vitro and in vivo as potent as other fly cell death proteins. Our observations suggest that DmHtrA2 promotes cell death through a cleavage of DIAP1 in the vicinity of mitochondria, which may represent a prototype of mitochondrial cell death pathway in evolution

  10. Studi Model Benchmark Mcnp6 Dalam Perhitungan Reaktivitas Batang Kendali Htr-10

    OpenAIRE

    Jupiter S.Pane, Zuhair, Suwoto, Putranto Ilham Yazid

    2016-01-01

    STUDI MODEL BENCHMARK MCNP6 DALAM PERHITUNGAN REAKTIVITAS BATANG KENDALI HTR-10. Dalam operasi reaktor nuklir, sistem batang kendali memainkan peranan yang sangat penting karena didesain untuk mengendalikan reaktivitas teras dan memadamkan reaktor. Nilai reaktivitas batang kendali harus diprediksi secara akurat melalui eksperimen dan perhitungan. Makalah ini mendiskusikan model Benchmark dalam perhitungan reaktivitas batang kendali reaktor HTR-10. Perhitungan dikerjakan dengan program transpo...

  11. Preliminary ripple effect analysis for HTR 350MWt 4 modules construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. H.; Lee, K. Y.; Shin, Y. J. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    We propose quantitative analysis techniques for ripple effects such as the production inducement effect and employment inducement effect for HTR 350MWt x 4 module construction and operation ripple effect based on NOAK. It is known that APR1400 reactors export ripple effect is about 8,500 billion KRW. As a result, HTR construction has more effective effect than that of APR1400.

  12. Vagal innervation is required for pulmonary function phenotype in Htr4-/- mice.

    Science.gov (United States)

    House, John S; Nichols, Cody E; Li, Huiling; Brandenberger, Christina; Virgincar, Rohan S; DeGraff, Laura M; Driehuys, Bastiaan; Zeldin, Darryl C; London, Stephanie J

    2017-04-01

    Human genome-wide association studies have identified over 50 loci associated with pulmonary function and related phenotypes, yet follow-up studies to determine causal genes or variants are rare. Single nucleotide polymorphisms in serotonin receptor 4 ( HTR4 ) are associated with human pulmonary function in genome-wide association studies and follow-up animal work has demonstrated that Htr4 is causally associated with pulmonary function in mice, although the precise mechanisms were not identified. We sought to elucidate the role of neural innervation and pulmonary architecture in the lung phenotype of Htr4 -/- animals. We report here that the Htr4 -/- phenotype in mouse is dependent on vagal innervation to the lung. Both ex vivo tracheal ring reactivity and in vivo flexiVent pulmonary functional analyses demonstrate that vagotomy abrogates the Htr4 -/- airway hyperresponsiveness phenotype. Hyperpolarized 3 He gas magnetic resonance imaging and stereological assessment of wild-type and Htr4 -/- mice reveal no observable differences in lung volume, inflation characteristics, or pulmonary microarchitecture. Finally, control of breathing experiments reveal substantive differences in baseline breathing characteristics between mice with/without functional HTR4 in breathing frequency, relaxation time, flow rate, minute volume, time of inspiration and expiration and breathing pauses. These results suggest that HTR4's role in pulmonary function likely relates to neural innervation and control of breathing. Copyright © 2017 the American Physiological Society.

  13. Knee Replacement

    Science.gov (United States)

    Knee replacement is surgery for people with severe knee damage. Knee replacement can relieve pain and allow you to ... Your doctor may recommend it if you have knee pain and medicine and other treatments are not ...

  14. Development and Reliability Analysis of HTR-PM Reactor Protection System

    International Nuclear Information System (INIS)

    Li Duo; Guo Chao; Xiong Huasheng

    2014-01-01

    High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) digital Reactor Protection System (RPS) is a dedicated system, which is designed and developed according to HTR-PM NPP protection specifications. To decrease the probability of accident trips and increase the system reliability, HTR-PM RPS has such features as a framework of four redundant channels, two diverse sub-systems in each channel, and two level two-out-of-four logic voters. Reliability analysis of HTR-PM RPS is based on fault tree model. A fault tree is built based on HTR-PM RPS Failure Modes and Effects Analysis (FMEA), and special analysis is focused on the sub-tree of redundant channel ''2-out-of-4'' logic and the fault tree under one channel is bypassed. The qualitative analysis of fault tree, such as RPS weakness according to minimal cut sets, is summarized in the paper. (author)

  15. Serotonergic gene polymorphisms (5-HTTLPR, 5HTR1A, 5HTR2A), and population differences in aggression: traditional (Hadza and Datoga) and industrial (Russians) populations compared.

    Science.gov (United States)

    Butovskaya, Marina L; Butovskaya, Polina R; Vasilyev, Vasiliy A; Sukhodolskaya, Jane M; Fekhredtinova, Dania I; Karelin, Dmitri V; Fedenok, Julia N; Mabulla, Audax Z P; Ryskov, Alexey P; Lazebny, Oleg E

    2018-04-16

    Current knowledge on genetic basis of aggressive behavior is still contradictory. This may be due to the fact that the majority of studies targeting associations between candidate genes and aggression are conducted on industrial societies and mainly dealing with various types of psychopathology and disorders. Because of that, our study was carried on healthy adult individuals of both sex (n = 853). Three populations were examined: two traditional (Hadza and Datoga) and one industrial (Russians), and the association of aggression with the following polymorphisms 5-HTTLPR, rs6295 (5HTR1A gene), and rs6311 (5HTR2A gene) were tested. Aggression was measured as total self-ratings on Buss-Perry Aggression Questionnaire. Distributions of allelic frequencies of 5-HTTLPR and 5HTR1A polymorphisms were significantly different among the three populations. Consequently, the association analyses for these two candidate genes were carried out separately for each population, while for the 5HTR2A polymorphism, it was conducted on the pooled data that made possible to introduce ethnic factor in the ANOVA model. The traditional biometrical approach revealed no sex differences in total aggression in all three samples. The three-way ANOVA (μ + 5-HTTLPR + 5HTR1A + 5HTR2A +ε) with measures of self-reported total aggression as dependent variable revealed significant effect of the second serotonin receptor gene polymorphism for the Hadza sample. For the Datoga, the interaction effect between 5-HTTLPR and 5HTR1A was significant. No significant effects of the used polymorphisms were obtained for Russians. The results of two-way ANOVA with ethnicity and the 5HTR2A polymorphism as main effects and their interactions revealed the highly significant effect of ethnicity, 5HTR2A polymorphism, and their interaction on total aggression. Our data provided obvious confirmation for the necessity to consider the population origin, as well as cultural background of tested individuals, while

  16. Potential of thorium use in the HTR reactor

    International Nuclear Information System (INIS)

    Engelmann, P.; Hansen, U.; Kolb, G.; Leushacke, D.; Teuchert, E.; Werner, H.

    1979-08-01

    In this investigation, several types of reactors and fuel circulations are dealt with as they refer to the region of the Federal Republic of Germany and are compared with each other as to their need for uranium and their costs until 2100. This includes also an investigation covering the effects of a postponed application of uranium-saving reactors, a delayed reprocessing and two variants of the nuclear energy's contribution to electricity generation. After today's light water reactor (LWR) of the pressure water reactor type (DWR) and the sodium-cooled fast breeder (SBR) which is being developed, the technically rather developed helium-cooled high temperature reactor (HTR) is dealt with as another system. The high temperature reactor is, because of its high coolant temperatures, not only suitable as a nuclear power plant, but can also be used to substitute fossile energy sources on the heat market and is being developed in Germany also for use as process heat reactor for nuclear coal gasification. Here the application of nuclear energy is only considered with regard to the region of power generation. Besides the case of the LWR and HTR-operation without reprocessing and fuel recycling for all reactor systems, the calculations also take into consideration the case of the closed fuel recycling. While LWR and SBR are based on the uranium-plutonium-fuel recycling, the thorium-uranium fuel circulation is considered for the HTR with globular fuel elements. As investigations made until today are generally restricted to the system LWR/SBR and the uranium-plutonium circulation, a main concern of the investigations presented here is to show the potential of the Thorium-utilization in high-temperature reactors and to determine how this system can also be applied during the time period concerned to set up a nuclear energy strategy which is safe and profitable as far as the uranium supply is concerned. (orig./UA) 891 UA/orig.- 892 HIS [de

  17. Does the HTR module have a chance for the future?

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1989-01-01

    The HTR module was developed as a robust and market-orientated heat source for a wide spectrum of applications. Its technology is largely based on that of the AVR. The choice of a low power density and the small core geometry permit thorough use to be made of the favourable safety characteristics and give an extra-ordinarily high degree of passive safety. There are possibilities for its introduction into the international market at present, particularly in the USSR and the People's Republic of China. (orig.)

  18. Prospective studies of HTR fuel cycles involving plutonium

    International Nuclear Information System (INIS)

    Bonin, B.; Greneche, D.; Carre, F.; Damian, F.; Doriath, J.Y.

    2002-01-01

    High Temperature Gas Cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry, and the parameters which characterize neutronic optimisation (moderation ratio or heavy nuclide concentration and distribution). Among other advantageous features, an HTR core has a better neutron economy than a LWR because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water) and in internal structures. Moreover, thanks to the high resistance of the coated particles, HTR fuels are able to reach very high burn-ups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors). These features make HTRs especially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory. A large number of fuel cycle studies are already available today, on 3 main categories of fuel cycles involving HTRs : i) High enriched uranium cycle, based on thorium utilization as a fertile material and HEU as a fissile material; ii) Low enriched uranium cycle, where only LEU is used (from 5% to 12%); iii) Plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials. Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. In the following, we complete the picture by a core study for a HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of fuel. With these limits in mind, we study in some detail the Pu cycle in the special case of a

  19. Research on application of burnable poison in pebble bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

    Burnable poison in fuel ball was used in pebble bed high-temperature gas-cooled reactor (HTR) to optimize the shape and the peak factor of power distribution in certain conditions. Two options are available and evaluated, that is the homogeneous burnable poison in graphite matrix and burnable poison particles (BPPs) in fuel balls. Due to the absorption cross section of "1"0B, the depletion speed for homogeneous burnable poison is very fast, and difficult to control, on the other side, the depletion speed of BPPs can be optimized respecting to its size, and better shape and peak value of power distribution can be achieved. (authors)

  20. Proceedings of the workshop on structural design criteria for HTR

    International Nuclear Information System (INIS)

    Breitbach, G.; Schubert, F.; Nickel, H.

    1989-04-01

    The papers demonstrate the status of high temperature reactor technology with regard to its realization in the nuclear power industry of various countries (FRG, USA, Japan) as well as to the development of safety rules in Germany. The design criteria for HTR could be presented. The criteria already determine definitely and almost completely the relevant requirements of the component rules. The informations include the technical boundary conditions with regard to safety, the metallic high temperature components, a particular section dealing with the reactor pressure vessel, especially with the prestressed concrete vessel, and the structural graphite components. (DG)

  1. First Results for Fluid Dynamics, Neutronics and Fission Product Behaviour in HTR applying the HTR Code Package (HCP) Prototype

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Kasselmann, S.; Xhonneux, A.; Lambertz, D.

    2014-01-01

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT are fed back into a new spectrum code of the HCP. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT–3D. Comparisons will be shown against data generated by the legacy codes VSOP99/11, NAKURE and FRESCO-II. (author)

  2. First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J., E-mail: h.j.allelein@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH Aachen University, 52064 Aachen (Germany); Kasselmann, S.; Xhonneux, A.; Tantillo, F.; Trabadela, A.; Lambertz, D. [Forschungszentrum Jülich, 52425 Jülich (Germany)

    2016-09-15

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

  3. DNA Methylation Analysis of HTR2A Regulatory Region in Leukocytes of Autistic Subjects.

    Science.gov (United States)

    Hranilovic, Dubravka; Blazevic, Sofia; Stefulj, Jasminka; Zill, Peter

    2016-02-01

    Disturbed brain and peripheral serotonin homeostasis is often found in subjects with autism spectrum disorder (ASD). The role of the serotonin receptor 2A (HTR2A) in the regulation of central and peripheral serotonin homeostasis, as well as its altered expression in autistic subjects, have implicated the HTR2A gene as a major candidate for the serotonin disturbance seen in autism. Several studies, yielding so far inconclusive results, have attempted to associate autism with a functional SNP -1438 G/A (rs6311) in the HTR2A promoter region, while possible contribution of epigenetic mechanisms, such as DNA methylation, to HTR2A dysregulation in autism has not yet been investigated. In this study, we compared the mean DNA methylation within the regulatory region of the HTR2A gene between autistic and control subjects. DNA methylation was analysed in peripheral blood leukocytes using bisulfite conversion and sequencing of the HTR2A region containing rs6311 polymorphism. Autistic subjects of rs6311 AG genotype displayed higher mean methylation levels within the analysed region than the corresponding controls (P epigenetic mechanisms might contribute to HTR2A dysregulation observed in individuals with ASD. © 2015 International Society for Autism Research, Wiley Periodicals, Inc.

  4. Preparation of spherical fuel elements for HTR-PM in INET

    International Nuclear Information System (INIS)

    Xiangwen, Zhou; Zhenming, Lu; Jie, Zhang; Bing, Liu; Yanwen, Zou; Chunhe, Tang; Yaping, Tang

    2013-01-01

    Highlights: • Modifications and optimizations in the manufacture of spherical fuel elements (SFE) for HTR-PM are presented. • A newly developed overcoater exhibits good stability and high efficiency in the preparation of overcoated particles. • The optimized carbonization process reduces the process time from 70 h in the period of HTR-10 to 20 h. • Properties of the prepared SFE and matrix graphite balls meet the design specifications for HTR-PM. • In particular the mean free uranium fraction of 5 consecutive batches is only 8.7 × 10 −6 . -- Abstract: The spherical fuel elements were successfully manufactured in the period of HTR-10. In order to satisfy the mass production of fuel elements for HTR-PM, several measures have been taken in modifying and optimizing the manufacture process of fuel elements. The newly developed overcoater system and its corresponding parameters exhibited good stability and high efficiency in the preparation of overcoated particles. The optimized carbonization process could reduce the carbonization time from more than 70 h to 20 h and improve the manufacturing efficiency. Properties of the manufactured spherical fuel elements and matrix graphite balls met the design specifications for HTR-PM. The mean free uranium fraction of 5 consecutive batches was 8.7 × 10 −6 . The optimized fuel elements manufacturing process could meet the requirements of design specifications of spherical fuel elements for HTR-PM

  5. Genetic variation in HTR4 and lung function: GWAS follow-up in mouse.

    Science.gov (United States)

    House, John S; Li, Huiling; DeGraff, Laura M; Flake, Gordon; Zeldin, Darryl C; London, Stephanie J

    2015-01-01

    Human genome-wide association studies (GWASs) have identified numerous associations between single nucleotide polymorphisms (SNPs) and pulmonary function. Proving that there is a causal relationship between GWAS SNPs, many of which are noncoding and without known functional impact, and these traits has been elusive. Furthermore, noncoding GWAS-identified SNPs may exert trans-regulatory effects rather than impact the proximal gene. Noncoding variants in 5-hydroxytryptamine (serotonin) receptor 4 (HTR4) are associated with pulmonary function in human GWASs. To gain insight into whether this association is causal, we tested whether Htr4-null mice have altered pulmonary function. We found that HTR4-deficient mice have 12% higher baseline lung resistance and also increased methacholine-induced airway hyperresponsiveness (AHR) as measured by lung resistance (27%), tissue resistance (48%), and tissue elastance (30%). Furthermore, Htr4-null mice were more sensitive to serotonin-induced AHR. In models of exposure to bacterial lipopolysaccharide, bleomycin, and allergic airway inflammation induced by house dust mites, pulmonary function and cytokine profiles in Htr4-null mice differed little from their wild-type controls. The findings of altered baseline lung function and increased AHR in Htr4-null mice support a causal relationship between genetic variation in HTR4 and pulmonary function identified in human GWAS. © FASEB.

  6. Survey of HTR related research at IRI, Delft, Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Wallerbos, E.J.M.; Van der Hagen, T.H.J.J.; Van Dam, H. [Interfaculty Reactor Institute IRI, Delft University of Technology, Delft (Netherlands); Tuerkcan, E. [ECN Nuclear Research, Petten (Netherlands)

    1998-09-01

    High temperature helium-cooled reactors have a large potential for inherent safety. Therefore, several projects on HTR research are being carried out or were carried out at the Interfaculty Reactor Institute (IRI) of the Delft University of Technology in Delft, Netherlands. As part of a larger research programme measurements of core reactivity, reactivity worth of safety rods and of small samples being oscillated in the reactor core were carried out at the PROTEUS facility of the Paul Scherrer Institute at Villigen, Switzerland. Together with other partners in the Netherlands a small inherently safe co-generation plant with a pebble-bed HTR core was designed and analysed. It was verified that such a reactor can operate continuously for 10 years by adding continuously fuel pebbles until the maximum available core height is reached. As a new, innovative, inherently safe reactor type the design of a fluidized-bed reactor with coated fuel particles on a helium gas stream is discussed and results are shown for the analysis of inherent criticality safety under varying coolant flow rates. IRI is also taking part in the new IAEA Co-ordinated Research Programme, which involves participation in the start-up experiments of the Japanese HTTR and carrying out calculations for the core physics benchmark test. 11 refs.

  7. HTR System Integration in Europe and South Africa

    International Nuclear Information System (INIS)

    Roelofs, Ferry; Ruer, J.; Cuadrado Garcia, P.; Cetnar, J.; Knoche, D.; Lapins, J.; Kasselman, S.; Stoker, P.; Fütterer, M.

    2014-01-01

    An HTR can be used for production of electricity and process heat. When these two applications are combined, a multitude of systems and components are needed. Whilst meeting the end user needs, this multitude of systems and components has to operate safely and economically. Therefore, within the framework of the European 7th framework program ARCHER project, a design schematic of a nuclear cogeneration system connected to a European and a South African industrial process is established and assessed. In order to provide an objective overview of the different indicators important for decision makers, the main characteristics with respect to the HTR system, the environment, safety, and economics are identified and compared to the characteristics of a modern gas turbine plant. In addition, a gap and SWOT analysis of a nuclear cogeneration system in Europe and South Africa are presented. In order to enable technical analysis of such a nuclear cogeneration system, a multitude of computer codes will be needed. Therefore, a code inventory is established of codes being used in Europe and South Africa for which the requirements for integration, development and qualification are assessed. (author)

  8. Core dynamics of HTR under ATWS and accident conditions

    International Nuclear Information System (INIS)

    Nabbi, R.

    1988-05-01

    The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

  9. Market prospects of modular HTR in EEC countries

    International Nuclear Information System (INIS)

    Albisu, F.; Garribba, S.F.; Lefevre, J.C.; Leuchs, D.; Vivante, C.

    1991-01-01

    The energy outlook for the early 21st century is very uncertain. Low-cost oil and natural gas reserves will become seriously depleted and nonfossil energy resources may be urgently required because of environmental reasons. In this framework, small- and medium-size nuclear reactors (SMSNRs), particularly the Modular High-Temperature Reactor (HTR) would allow extension of uses of nuclear energy while being adopted to produce power and/or steam or heat, where heat can be at low or high temperature. For policy making and planning purposes it is meaningful to appraise the market potential of Modular HTR during the next 20 or 30 years. The paper presents the outcomes of country studies on the subject conducted for a sample of EC Member nations, including France, Federal Republic of Germany, Italy, and Spain. Among the goals of the studies are the definition of market segments, and identification of the principal obstacles which will affect future adoption of SMSNRs. Opportunities offered by the different contexts and energy end-uses seem promising. Numerous difficulties and constraints emerge, however, some of which might be eased by actions that national governments or more often the European Community may wish to take. (author)

  10. Market prospects of modular HTR in EEC countries

    International Nuclear Information System (INIS)

    Albisu, F.; Garribba, S.F.; Lefevre, J.C.; Leuchs, D.; Vivante, C.

    1992-01-01

    The energy outlook for the early 21st century is very uncertain. Low-cost oil and natural gas reserves will become seriously depleted and non-fossil energy resources may be urgently required because of environmental reasons. In this framework, the European Economic Community should be able to rely upon nuclear energy as an economic, safe and readily deployable resource for its future. Small and medium-size nuclear reactors (SMSNRs), particularly modular high-temperature reactor (HTR) would allow extension of uses of nuclear energy while being adopted to produce power and/or steam or heat, where heat can be at low or high temperature. For policy making and planning purposes it appears meaningful to appraise the market potential of modular HTR during the next twenty or thirty years. Thus the paper presents the outcomes of country studies on the subject conducted for a sample of Member nations to the European Economic Community including France, Federal Republic of Germany, Italy and Spain. Amongst the goals of the studies are definition of market segments, identification of the principal obstacles which would affect future adoption of SMSNRs. Opportunities offered by the different contexts and energy end-uses seem promising. Numerous difficulties and constraints emerge however, some of which might be eased by actions that national governments or more often the European Economic Community, may wish to take. (orig.)

  11. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  12. The nucleus accumbens 5-HTR4-CART pathway ties anorexia to hyperactivity

    Science.gov (United States)

    Jean, A; Laurent, L; Bockaert, J; Charnay, Y; Dusticier, N; Nieoullon, A; Barrot, M; Neve, R; Compan, V

    2012-01-01

    In mental diseases, the brain does not systematically adjust motor activity to feeding. Probably, the most outlined example is the association between hyperactivity and anorexia in Anorexia nervosa. The neural underpinnings of this ‘paradox', however, are poorly elucidated. Although anorexia and hyperactivity prevail over self-preservation, both symptoms rarely exist independently, suggesting commonalities in neural pathways, most likely in the reward system. We previously discovered an addictive molecular facet of anorexia, involving production, in the nucleus accumbens (NAc), of the same transcripts stimulated in response to cocaine and amphetamine (CART) upon stimulation of the 5-HT4 receptors (5-HTR4) or MDMA (ecstasy). Here, we tested whether this pathway predisposes not only to anorexia but also to hyperactivity. Following food restriction, mice are expected to overeat. However, selecting hyperactive and addiction-related animal models, we observed that mice lacking 5-HTR1B self-imposed food restriction after deprivation and still displayed anorexia and hyperactivity after ecstasy. Decryption of the mechanisms showed a gain-of-function of 5-HTR4 in the absence of 5-HTR1B, associated with CART surplus in the NAc and not in other brain areas. NAc-5-HTR4 overexpression upregulated NAc-CART, provoked anorexia and hyperactivity. NAc-5-HTR4 knockdown or blockade reduced ecstasy-induced hyperactivity. Finally, NAc-CART knockdown suppressed hyperactivity upon stimulation of the NAc-5-HTR4. Additionally, inactivating NAc-5-HTR4 suppressed ecstasy's preference, strengthening the rewarding facet of anorexia. In conclusion, the NAc-5-HTR4/CART pathway establishes a ‘tight-junction' between anorexia and hyperactivity, suggesting the existence of a primary functional unit susceptible to limit overeating associated with resting following homeostasis rules. PMID:23233022

  13. The present state of the HTR concept based on experience gained from AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, W.

    1989-01-01

    During the past ten years the development of a specific HTR concept has made remarkable progress. This has been mainly characterized by making use of the safety characteristics typical of the High-Temperature Reactor (HTR). In the design, construction and operation of High-Temperature Reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - comprehensive experience has been gained in the field of operational availability and safety, accident topology and plant risk of HTRs in recent years. This experience is relevant for the entire HTR line independent of specific projects. (author). 3 refs, 5 figs, 1 tab

  14. Simulation of Thermal-hydraulic Process in Reactor of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

    2014-01-01

    This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

  15. Applications and Prospects of Modularization Technology in HTR Project Starting from Primary Loop Cavity Construction

    International Nuclear Information System (INIS)

    Yang Guokang; Chen Jing; Huang Wen; Lin Lizhi; Sun Yunlun; Chen Yan; Mao Jiaxin; Wang Yougang; Wang Jinwen; Lin Mingfeng; Yang Mingshan

    2014-01-01

    Primary loop cavity is one of the key areas and major difficulties in HTR-PM project construction. In order to shorten the construction schedule and improve the construction quality, researches on modular design and construction of primary loop cavity has been carried out and the results have been applied in HTR-PM project construction, and got significant application benefit. This paper summarizes the modularization technology application research and project implementation results of primary loop cavity, and analyzes the application and prospects of modularization technology in the HTR project construction. (author)

  16. Design of the steam reformer for the HTR-10 high temperature process heat application

    International Nuclear Information System (INIS)

    Ju Huaiming; Xu Yuanhui; Jia Haijun

    2000-01-01

    The 10 MW High Temperature Reactor Test Module (HTR-10) is being constructed now and planned to be operational in 2000. One of the objectives is to develop the high temperature process heat application. The methane steam reformer is one of the key-facilities for the nuclear process heat application system. The paper describes the conceptual design of the HTR-10 Steam Reformer with He heating, and the design optimization computer code. It can be used to perform sensitivity analysis for parameters, and to improve the design. Principal parameters and construction features of the HTR-10 reformer heated by He are introduced. (author)

  17. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  18. Burning minor actinides in a HTR energy spectrum

    International Nuclear Information System (INIS)

    Pohl, Christoph; Rütten, H. Jochem

    2012-01-01

    Highlights: ► Burn-up analysis for varying plutonium/minor actinide fuel compositions. ► The influence of varying heavy metal fuel element loads is investigated. ► Significant burn-up via radiative capture and subsequently fission is observed. ► Difference observed between fuel element burn-up and total actinide burning rate. - Abstract: The generation of nuclear energy by means of the existing nuclear reactor systems is based mainly on the fission of U-235. But this comes along with the capture of neutrons by the U-238 faction and results in a build-up of plutonium isotopes and minor actinides as neptunium, americium and curium. These actinides are dominant for the long time assessment of the radiological risk of a final disposal therefore a minimization of the long living isotopes is aspired. Burning the actinides in a high temperature helium cooled graphite moderated reactor (HTR) is one of these options. The use of plutonium isotopes to sustain the criticality of the system is intended to avoid on the one hand highly enriched uranium because of international regulations and on the other hand low enriched uranium because of the build up of new actinides from neutron capture in the U-238 fraction. Because initial minor actinide isotopes are typically not fissionable by thermal neutrons the idea is to fission instead the intermediate isotopes generated by the first neutron capture. This paper comprises calculations for plutonium/minor actinides/thorium fuel compositions and their correlated final burn-up for a generic pebble bed HTR based on the reference design of the 400 MW PBMR. In particular the cross sections and the neutron balance of the different minor actinide isotopes in the higher thermal energy spectrum of a HTR will be discussed. For a fuel mixture of plutonium and minor actinides a significant burn-up of these actinides up to 20% can be achieved but at the expense of a higher residual fraction of plutonium in the burned fuel. Combining

  19. The HTR 500 concept based on pratical THTR and AVR experience

    International Nuclear Information System (INIS)

    Wachholz, W.; Weicht, U.

    1988-01-01

    This paper discusses progress during the past ten years in the development of a specific HTR safety concept. This has been mainly characterized by the abandonment of the LWR specific safety principles and making use of the safety characteristics typical of the high-temperature reactor (HTR). In the design, construction and operation of high-temperature reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - experience has been gained in the field of accident topology and plant risk of HTRs in recent years. This experience, based on detailed accident analyses performed by manufacturers and experts, is relevant for the entire HTR line independent of specific projects. The authors focus on the HTR 500, the first commercial high temperature reactor with a pebble bed core. Its design principles and the design of its systems are based on the earlier AVR and THTR projects

  20. Study on the production mechanism of Co-60 in the primary loop of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Xie Feng; Li Hong; Cao Jianzhu; Li Fu; Wei Liqiang

    2015-01-01

    Co-60 is an activated metallic erosion product, which is very important for waste management and decommissioning work of pressurized water reactor (PWR) power plants. Recent measurement on the samples from the primary loop of HTR-10 indicates the existence of Co-60. In current paper, the preliminary experimental results in HTR-10 will be introduced, and the production mechanism of Co-60 in the pebble bed high temperature gas-cooled reactors will be summarized and compared with that in PWRs and Germany High Temperature Nuclear Reactor (AVR). The further experiments with decomposing the post-irradiation graphite spheres of HTR-10 are put forward, which will promote the further study to testify the production sources of Co-60 and be of great significance in the waste minimization and the decommissioning work of HTR-10. (author)

  1. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  2. Numerical Simulation of Two-branch Hot Gas Mixing at Reactor Outlet of HTR-PM

    International Nuclear Information System (INIS)

    Hao Pengefei; Zhou Yangping; Li Fu; Shi Lei; He Heng

    2014-01-01

    A series of two-branch model experiment has been finished to investigate the thermal mixing efficiency of the HTR-PM reactor outlet. This paper introduces the numerical simulation on the design of thermal mixing structure of HTR-PM and the test facility with Fluent software. The profiles of temperature, pressure and velocity in the mixing structure design and the test facility are discussed by comparing with the model experiment results. The numerical simulation results of the test facility have good agreement to the experiment results. In addition, the thermal-fluid characters obtained by numerical simulation show the thermal mixing structure of HTR-PM has similarity with the test facility. Finally, it is concluded that the thermal mixing design at HTR-PM reactor outlet can fulfilled the requirements for high thermal mixing efficiency and appropriate pressure drop. (author)

  3. Experiment study on thermal mixing performance of HTR-PM reactor outlet

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-09-15

    A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

  4. Replacing penalties

    Directory of Open Access Journals (Sweden)

    Vitaly Stepashin

    2017-01-01

    Full Text Available УДК 343.24The subject. The article deals with the problem of the use of "substitute" penalties.The purpose of the article is to identify criminal and legal criteria for: selecting the replacement punishment; proportionality replacement leave punishment to others (the formalization of replacement; actually increasing the punishment (worsening of legal situation of the convicted.Methodology.The author uses the method of analysis and synthesis, formal legal method.Results. Replacing the punishment more severe as a result of malicious evasion from serving accused designated penalty requires the optimization of the following areas: 1 the selection of a substitute punishment; 2 replacement of proportionality is serving a sentence other (formalization of replacement; 3 ensuring the actual toughening penalties (deterioration of the legal status of the convict. It is important that the first two requirements pro-vide savings of repression in the implementation of the replacement of one form of punishment to others.Replacement of punishment on their own do not have any specifics. However, it is necessary to compare them with the contents of the punishment, which the convict from serving maliciously evaded. First, substitute the punishment should assume a more significant range of restrictions and deprivation of certain rights of the convict. Second, the perfor-mance characteristics of order substitute the punishment should assume guarantee imple-mentation of the new measures.With regard to replacing all forms of punishment are set significant limitations in the application that, in some cases, eliminates the possibility of replacement of the sentence, from serving where there has been willful evasion, a stricter measure of state coercion. It is important in the context of the topic and the possibility of a sentence of imprisonment as a substitute punishment in cases where the original purpose of the strict measures excluded. It is noteworthy that the

  5. Rework of process effluents from the fabrication of HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lasberg, Ingo; Braehler, Georg [NUKEM Technologies GmbH (Germany); Boyes, David [Pebble Bed Modular Reactor (Pty) Ltd., Centurion (South Africa)

    2008-07-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m{sup 3}/a), isopropanol IPA/water mixtures (130 m{sup 3}/a); Non-Process Water NPW (300 m{sup 3}/a); methanol (7m{sup 3}/a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  6. Rework of process effluents from the fabrication of HTR fuel

    International Nuclear Information System (INIS)

    Lasberg, Ingo; Braehler, Georg; Boyes, David

    2008-01-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m 3 /a), isopropanol IPA/water mixtures (130 m 3 /a); Non-Process Water NPW (300 m 3 /a); methanol (7m 3 /a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  7. The Hitrex Programme: unperturbed HTR lattice and control rod measurements

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A J; Nunn, D L

    1972-06-15

    Reactivity, power distributions, plutonium production and fast neutron graphite damage are being studied at Berkeley Nuclear Laboratories (BNL) on the HTR 'Hitrex' reactor under cold clean conditions. Rod interactions, important in assessing local criticality hazards, are receiving special attention in the measurements. The proposals for the first two series of measurements on Hitrex are discussed in this note, Hitrex 1a being the unperturbed reactor, and Hitrex 1b the same fuel array but with a number of different control absorber loadings in it. Common to both series will be cross pin, cross block and cross core measurements of power rating, thermal spectrum and damage dose distributions, so that these will be known as functions of the fuel, reflector and absorber environment.

  8. HTR-500 - a technical and engineered safeguards concept

    International Nuclear Information System (INIS)

    Schoening, J.; Wachholz, W.; Stoelzl, D.

    1985-01-01

    The plant succeeding the THTR-300 nuclear power plant, which has just started its trial phase of power operation, is the HTR-500. On behalf of the Arbeitsgemeinschaft Hochtemperaturreaktor (AHR), the BBC/HRB Group completed a preliminary project study of a nuclear power plant equipped with a high temperature reactor in the 500 MW power range, in which the changed requirements in the nuclear power market are taken into account and electricity generating costs are to be achieved which are competitive with those of a 1230 MW convoy pressurized water reactor of the present design. On this basis, construction documents are to be drafted, and the licensing procedure under the Atomic Energy Act is to be carried out, within a planning phase of roughly four years. (orig.) [de

  9. 'Once through' cycles in the pebble bed HTR

    International Nuclear Information System (INIS)

    Teuchert, E.

    1977-12-01

    In the pebble bed HTR the 'Once Through' cycles achieve a favorable conservation of uranium resources due to their high burnup and due to the relatively low fissile inventory. A detailed study is given for cycles with highly enriched uranium and thorium, 20% enriched uranium and thorium, and for the low (approximately 8%) enriched cycle. The recommended cycle is based on the known THTR fuel element in the Th/U (93%) cycle. The variant with separate Seed elements and Breed elements presents the best pioneer in view of later recycling and thermal breeding. The minimum proliferation risk is achieved in the Th/U (20%) cycle basing on the fuel element type of the AVR, due to the low amount and high denaturization of the disloaded plutonium. (orig.) [de

  10. Different contributions of HtrA protease and chaperone activities to Campylobacter jejuni stress tolerance and physiology

    DEFF Research Database (Denmark)

    Bæk, Kristoffer Torbjørn; Vegge, Christina Skovgaard; Skórko-Glonek, Joanna

    2011-01-01

    activity is sufficient for growth at high temperature or oxidative stress, whereas the HtrA protease activity is only essential at conditions close to the growth limit for C. jejuni. However, the protease activity was required to prevent induction of the cytoplasmic heat-shock response even at optimal......The microaerophilic bacterium Campylobacter jejuni is the most common cause of bacterial food-borne infections in the developed world. Tolerance to environmental stress relies on proteases and chaperones in the cell envelope such as HtrA and SurA. HtrA displays both chaperone and protease activity......, but little is known about how each of these activities contributes to stress tolerance in bacteria. In vitro experiments showed temperature dependent protease and chaperone activities of C. jejuni HtrA. A C. jejuni mutant lacking only the protease activity of HtrA was used to show that the HtrA chaperone...

  11. Relationship between the Toyo Tanso Group and HTR-PM

    International Nuclear Information System (INIS)

    Zhan Guobin; Konishi, Takashi

    2014-01-01

    IG-110 that is Isotropic graphite for nuclear applications, is the only product that is used for two types of High Temperature Gas-cooled Reactors, prismatic type HTTR and pebble-bed type HTR-10, that are currently in operation in the world. IG-110 is highly evaluated in the global market for its track record and physical stability. The Toyo Tanso Group won the contract to build graphite core internals for HTR-PM that is a world’s first modular pebble-bed high temperature gas-cooled demonstration reactor. A decision was made to manufacture IG-110 graphite materials at Toyo Tanso Japan called TTJ and to process products and undertake temporary assembly at Shanghai Toyo Tanso called STT. Manufacture of graphite materials for which TTJ is responsible has been completed. As the next step, processing of products is scheduled to commence at STT from this autumn. Our graphite materials were required to be 2,000 mm or more in maximum length. The number of graphite blocks required exceeded 3,500. Although the graphite structure requirements including configuration were highly challenging, we were able to meet all the requirements with our engineering capabilities, i.e. decades of track record in manufacture and stability in characteristics. STT that will start the machining process this autumn is equipped with state-of-the-art processing machines and three-dimensional measuring machines. Notably, STT has high levels of engineering capabilities to process and inspect tens of thousands of internal components for reactors in accordance with drawings and to temporarily assemble these components. (author)

  12. Factors influencing selection of a HTR for a developing country

    International Nuclear Information System (INIS)

    Karim, C.S.

    1989-01-01

    Consumption of commercial energy and electricity in Bangladesh has to grow rapidly in order to attain socio-economic development of the country. Nuclear power is considered to be an appropriate proposition due to the inadequacy of indigenous primary energy resources. A technical, economic and financial feasibility study of a 300-500 MWe nuclear power plant is underway now. Responses from different suppliers in SMPR range were enumerated jointly by the Consultants and BAEC under the feasibility study. Criteria for selection of technology and the factor influencing the selection of Modular HTR for Bangladesh are described in the paper. Some indicative results of cost economic calculations are included to help form an idea about various limiting conditions, under which a SMPR with the selected technology could become competitive with the other conventional alternatives. Problems in decision making associated with the uncertainties in estimating plant and fuel cycle costs are enumerated. The implications of not having a reference plant vis-a-vis the advantageous safety features are described to show how these aspects can influence the selection of a new technology like HTR for a developing country. Financing is identifiable as the major problem in implementing a nuclear power project in a developing country like Bangladesh. The entire scope of supplies and services may be broken down into components, so that the burden of financing could be shared by more than one exporting country. Some indicative ideas about the packaging of supplies and services are presented in the paper in order to identify different types of financing sources that could be explored for implementation of the project. Some salient features of the effect of joint-venture on the project financing and implementation are described in the paper. (author). 3 refs, 1 fig

  13. Development of digital I&C system in HTR-PM

    International Nuclear Information System (INIS)

    Shi Guilian

    2014-01-01

    Conclusions: HTR-PM DCS has been under execution for 5 years( 2009-2014) . It has taken CTEC 150 man/year so far. With close cooperation with INET, Chinergyand Shanghai Electric, CTEC overcame difficulties, like iterative design, voluminous customization work, new technology, and lacking of drawings. However, the accomplishment of the planned milestones prepared CTEC for the following work in HTR-PM DCS. 1. The 1ST integrated DCS, including safety DCS, non-safety DCS, DEH supplied by Chinese supplier. Rod control system and DEH are integrated in non-safety DCS. Simplified interface, integrated platform, and easy to use and maintenance. 2. CTEC obtained knowledge of 4th generation HTR-PM digital I&C, key design technology, and riched its DCS products by participation in HTRPM. HTR-PM Safety DCS project provided valuable experience for CTEC’s development and application of FIRMSYS, a safety protection control system platform. 3. The qualification solution by customized HTR-PM safety DCS prototype helps simply safety DCS design, V&V, qualification and safety review of the actual system, but results in some problems in system upgrade and maintenance. With the satisfactory application of FIRMSYS in 1000mw PWR and platform qualification , the future HTR-PM safety DCS could be provided based on a qualified safety DCS platform.

  14. Design of reactor protection systems for HTR plants generating electric power and process heat problems and solutions

    International Nuclear Information System (INIS)

    Craemer, B.; Dahm, H.; Spillekothen, H.G.

    1982-06-01

    The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)

  15. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  16. Current status and technical description of Chinese 2 x 250 MWth HTR-PM demonstration plant

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wu Zongxin; Wang Dazhong; Xu Yuanhui; Sun Yuliang; Li Fu; Dong Yujie

    2009-01-01

    After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10. Due to intensive engineering and R and D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled. The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW. A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line. The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures. According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started. After the successful operation of the demonstration plant

  17. Spent sulfite liquor developments

    Energy Technology Data Exchange (ETDEWEB)

    Black, H H

    1958-01-01

    A review of methods of utilizing spent sulfite liquor, including evaporation and burning, fermentation to produce yeast or alcohol, production of vanillin and lignosulfonates, and use as a roadbinder.

  18. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    Poinssot, Ch.

    2002-01-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

  19. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  20. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

    International Nuclear Information System (INIS)

    1979-08-01

    This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

  1. Knee Replacement

    Science.gov (United States)

    ... days. Medications prescribed by your doctor should help control pain. During the hospital stay, you'll be encouraged to move your ... exercise your new knee. After you leave the hospital, you'll continue physical ... mobility and a better quality of life. And most knee replacements can be ...

  2. Proceedings of the Fifth Seminar of High Temperature Reactor: The Role and Challenge with HTR Opportunity in the Twenty-first Century

    International Nuclear Information System (INIS)

    As-Natio-Lasman; Zaki-Su'ud; Bambang-Sugiono

    2000-11-01

    The Seminar in HTR Reactor has become routine activities held in BATAN since 1994. This Seminar is a continuation of the Seminar on Technology and HTR Application held by Centre for Development of Advanced Reactor System. The theme of the seminar is Role, Challenge, Opportunity of HTR in the Twenty-first Century. Thirteen papers presented in the seminar were collected into proceedings. The aims of the proceedings is to provide information and references on nuclear technology, mainly on HTR technology. (DII)

  3. Spent fuel storage facility, Kalpakkam

    International Nuclear Information System (INIS)

    Shreekumar, B.; Anthony, S.

    2017-01-01

    Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements

  4. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

    Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  5. Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

    International Nuclear Information System (INIS)

    Sun Yanfei; Zhong Shuoping; Huang Xiaojin

    2014-01-01

    Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

  6. The Research Status for Decommissioning and Radioactive Waste Minimization of HTR-PM

    International Nuclear Information System (INIS)

    Li Wenqian; Li Hong; Cao Jianzhu; Tong Jiejuan

    2014-01-01

    Decommissioning of the high-temperature gas-cooled reactor-pebble bed module (HTR-PM) as a part of the nuclear power plant, is very important during the early design stage of the construction, and it is under study and research currently. This article gives a thorough description of the current decommissioning study status of HTR-PM. Since HTR-PM has its features such as adopting a large amount of graphite, the waste inventory and characterization will be quite different from other type of reactors, new researches should be carried out and good lessons of practices and experiences should be learned from international other reactors, especially the AVR. Based on the new international regulations and Chinese laws, a comprehensive decommissioning program should be proposed to guarantee the HTR-PM will succeed in every stage of the decommissioning, such as defueling, decontamination, dismantling, demolition, waste classification and disposal, etc. In the meantime, the minimization of the radioactive waste should be taken into account during the whole process - before construction, during operation and after shut down. In this article, the decommissioning strategy and program conception of HTR-PM will be introduced, the radiation protection consideration during the decommissioning activities will be discussed, and the research on the activation problem of the decommissioning graphite will be introduced. (author)

  7. Design and Experiment of Auxiliary Bearing for Helium Blower of HTR-PM

    International Nuclear Information System (INIS)

    Yang Guojun; Shi Zhengang; Liu Xingnan; Zhao Jingjing

    2014-01-01

    The helium blower is the important equipment for HTR-PM. Active magnetic bearing (AMB) instead of mechanical bearing is selected to support the rotor of the helium blower. However, one implication of AMB is the requirement to provide the auxiliary bearing to mitigate the effects of failures or overload conditions. The auxiliary bearing is used to support the rotor when the AMB fails to work. It must support the dropping rotor and bear the great impact force and friction heat. The design of the auxiliary bearing is one of the challenging problems in the whole system. It is very important for the helium blower with AMB of HTR-PM to make success. The rotor’s length of helium blower of HTR-PM is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The axial load is 4500kg, and the radial load is 1950kg. The angular contact ball bearing was selected as the auxiliary bearing. The test rig has been finished. It is difficult to analyze the falling course of the rotor. The preliminary analysis of the dropping rotor was done in the special condition. The impact force of auxiliary bearing was computed for the axial and radial load. And the dropping test of the blower rotor for HTR-10 will be introduced also in this paper. Results offer the important theoretical base for the protector design of the helium blower with AMB for HTR-PM. (author)

  8. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  9. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  10. Disposal of spent fuel

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

    1978-01-01

    Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

  11. HTR1B as a risk profile maker in psychiatric disorders: a review through motivation and memory.

    Science.gov (United States)

    Drago, Antonio; Alboni, Silvia; Brunello, Nicoletta; Nicoletta, Brunello; De Ronchi, Diana; Serretti, Alessandro

    2010-01-01

    Serotonin receptor 1B (HTR1B) is involved in the regulation of the serotonin system, playing different roles in specific areas of the brain. We review the characteristics of the gene coding for HTR1B, its product and the functional role of HTR1B in the neural networks involved in motivation and memory; the central role played by HTR1B in these functions is thoroughly depicted and show HTR1B to be a candidate modulator of the mnemonic and motivationally related symptoms in psychiatric illnesses. In order to challenge this assessment, we analyze how and how much the genetic variations located in the gene that codes for HTR1B impacts on the psychiatric phenotypes by reviewing the literature on this topic. We gathered partial evidence arising from genetic association studies, which suggests that HTR1B plays a relevant role in substance-related and obsessive compulsive disorders. On the other hand, no solid evidence for other psychiatric disorders was found. This finding is quite striking because of the heavy impairment of motivation and of mnemonic-related functions (for example, recall bias) that characterize major psychiatric disorders. The possible reasons for the contrast between the prime relevance of HTR1B in regulating memory and motivation and the limited evidence brought by genetic association studies in humans are discussed, and some suggestions for possible future directions are provided.

  12. Fission product behaviour in the primary circuit of an HTR

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Iniotakis, N.

    1981-01-01

    The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

  13. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland; Köberl, Oliver

    2014-01-01

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  14. Design Procedure of Graphite Components by ASME HTR Codes

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Jo, Chang Keun

    2016-01-01

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

  15. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

  16. Axial temperatures and fuel management models for a HTR system

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-11-12

    In the HTR system, there is a large difference in temperature between different parts of the reactor core. The softer neutron spectrum in the upper colder core regions tends to shift the power productions in the fresh fuel upwards. As uranium 235 depletes and plutonium with its higher cross sections in the lower hot regions is built-up, an axial power flattening takes place. These effects have been studied in detail for a single column in an equilibrium environment. The aim of this paper is to relate these findings to a whole reactor core and to investigate the influence of axial temperatures on the overall performance and in particular, the fuel management scheme chosen for the reference design. A further objective has been to calculate the reactivity requirements for different part load conditions and for various daily and weekly load diagrams. As the xenon cross section changes significantly with temperature these investigations are performed for an equilibrium core with due representation of axial temperature zones.

  17. Burnup measurement study and prototype development in HTR-PM

    International Nuclear Information System (INIS)

    Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

    2014-01-01

    In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)

  18. The behaviour of spherical HTR fuel elements under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

  19. Impact of the Improved Resonance Scattering Kernel on HTR Calculations

    International Nuclear Information System (INIS)

    Becker, B.; Dagan, R.; Broeders, C.H.M.; Lohnert, G.

    2008-01-01

    The importance of an advanced neutron scattering model for heavy isotopes with strong energy dependent cross sections such as the pronounced resonances of U 238 has been discussed in various publications where the full double differential scattering kernel was derived. In this study we quantify the effect of the new scattering model for specific innovative types of High Temperature Reactor (HTR) systems which commonly exhibit a higher degree of heterogeneity and higher fuel temperatures, hence increasing the importance of the secondary neutron energy distribution. In particular the impact on the multiplication factor (k ∞ ) and the Doppler reactivity coefficient is presented in view of the packing factors and operating temperatures. A considerable reduction of k ∞ (up to 600 pcm) and an increased Doppler reactivity (up to 10%) is observed. An increase of up to 2.3% of the Pu 239 inventory can be noticed at 90 MWd/tHM burnup due to enhanced neutron absorption of U 238 . Those effects are more pronounced for design cases in which the neutron flux spectrum is hardened towards the resolved resonance range. (authors)

  20. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  1. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    McFarlane, L.F.; Lineberry, M.J.

    1995-01-01

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  2. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    López Lizana, F.

    2015-01-01

    The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)

  3. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

  4. Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

    International Nuclear Information System (INIS)

    Sun Yunxue; Shao Jin

    2015-01-01

    Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

  5. Analysis the Response Function of the HTR Ex-core Neutron Detectors in Different Core Status

    International Nuclear Information System (INIS)

    Fan Kai; Li Fu; Zhou Xuhua

    2014-01-01

    Modular high temperature gas cooled reactor HTR-PM demonstration plant, designed by INET, Tsinghua University, is being built in Shidao Bay, Shandong province, China. HTR-PM adopts pebble bed concept. The harmonic synthesis method has been developed to reconstruct the power distributions on HTR-PM. The method based on the assumption that the neutron detector readings are mainly determined by the status of the core through the power distribution, and the response functions changed little when the status of the core changed. To verify the assumption, the influence factors to the ex-core neutron detectors are calculated in this paper, including the control rod position and the temperature of the core. The results shows that when the status of the core changed, the power distribution changed more remarkable than the response function, but the detector readings could change about 5% because of the response function changing. (author)

  6. Simulation and study on reactivity disturbs dynamic character of HTR-10 nuclear power system

    International Nuclear Information System (INIS)

    Huang Xiaojin; Feng Yuankun

    2002-01-01

    In order to not only know 10 MW High Temperature Gas Cooled Reactor (HTR-10) nuclear power system's dynamic character more deeply but also to satisfy requirements of control system's design and analysis, the dynamic model of HTR-10 nuclear power system is established on the basis of dynamic model of HTR-10 nuclear system, which supplies turbine and generate electricity system model. Using this model, system's main variables' dynamic processes are simulated when control rod takes step reactivity disturb. The concussive progresses which is caused by reactivity disturb are analyzed. The results indicate that fuel temperature changing more slowly than nuclear power makes reactivity negative feedback not to restrain power changing, and then power concussive progress comes to being

  7. Coal conversion and the HTR - basic elements of novel power supply concepts

    International Nuclear Information System (INIS)

    Buerger, F.H.

    1985-01-01

    A meeting under this title was held in Dortmund on 16 to 19 September, 1985, jointly by the VGB Technische Vereinigung der Grosskraftwerksbetreiber e.V., Essen, and the Vereinigte Elektrizitaetswerke Westfalen AG (VEW), Dortmund. The meeting was held in two sections: 'Gersteinwerk power plant - the combination unit K and the KUV coal conversion system' and '7th International conference on HTR technology'. Three technologies were discussed that will have a significant role on the future energy market, i.e., the HTR reactor line (first applied in the Hamm-Uentrop THTR reactor), the new generation of coal-fired power plants with combined gas/steam turbines, and the coal gasification technology. All three systems will make more efficient and less-polluting use of domestic coal by using HTR process heat, by converting coal to widen its range of applications, and by providing more efficient combination units for power plants. (orig./UA) [de

  8. Numerical analysis of magnetically suspended rotor in HTR-10 helium circulator being dropped into auxiliary bearings

    International Nuclear Information System (INIS)

    Zhao Jingxiong; Yang Guojun; Li Yue; Yu Suyuan

    2012-01-01

    Active magnetic bearings (AMB) have been selected to support the rotor of primary helium circulator in commercial 10 Mega-Walt High Temperature Gas-cooled Reactor (HTR-10). In an AMB system, the auxiliary bearings are necessary to protect the AMB components in case of losing power. This paper performs the impact simulation of Magnetically Suspended Rotor in HTR-10 Helium Circulator being dropped into the auxiliary bearings using the finite element program ABAQUS. The dynamic response and the strain field of auxiliary bearings are analyzed. The results achieved by the numerical analysis are in agreement with the experiment results. Therefore, the feasibility of the design of auxiliary bearing and the possibility of using the AMB system in the HTR are proved. (authors)

  9. Concept of a HTR modular plant for generation of process heat in a chemical plant

    International Nuclear Information System (INIS)

    1991-07-01

    This final report summarizes the results of a preliminary study on behalf of Buna AG and Leunawerke AG. With regard to the individual situations the study investigated the conditions for modular HTR-2 reactors to cover on-site process heat and electric power demands. HTR-2 reactor erection and operation were analyzed for their economic efficiency compared with fossil-fuel power plants. Considering the prospective product lines, the technical and economic conditions were developed in close cooperation with Buna AG and Leunawerke AG. The study focused on the technical integration of modular HTR reactors into plants with regard to safety concepts, on planning, acceptance and erection concepts which largely exclude uncalculable scheduling and financial risks, and on comparative economic analyses with regard to fossil-fuel power plants. (orig.) [de

  10. Actual characteristics study on HTR-10GT coupling with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Peng Xuechuang; Zhu Shutang; Wang Jie

    2005-01-01

    HTR-10GT is a testing project coupling the reactor HTR-10 with direct gas turbine cycle. Its thermal cycle can be taken as a closed, recuperated and inter-cooled Brayton cycle. The present study is focused on the thermal cycle performance of HTR-10GT under practical conditions of leakage, pressure losses, etc.. Through thermodynamic analysis, the expression of cycle efficiency for actual thermal cycle is derived. By establishing a physical model with friction loss and leakage, a set of governing equation are constructed based on some reasonable assumptions. The results of actual cycle efficiency have been calculated for different leakage amount at different locations while the effects of leakage under different power level have also been calculated and analyzed. (authors)

  11. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  12. Design and application of the HTR-100 industrial nuclear power plant

    International Nuclear Information System (INIS)

    Brandes, S.; Kohl, W.

    1988-01-01

    The small HTR-100 high temperature reactor combines the reactor concept of the AVR reactor, which has been proven for 20 years, with the latest component technology of the THTR power plant which has been in operation since 1985. The nuclear heat supply system is conceived so as to be applicable for the generation of electric power, district heat and process steam according to the customer's demand. The HTR-100 reactor has a thermal power of 258 MW and offers steam parameters of 190 bar/530 0 C. To cover a higher power demand HTR-100 reactors can be combined forming a larger power plant. Economic analyses have shown competitiveness with fossil power plants. (orig.)

  13. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

    This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

  14. Calculation of the Fission Product Release for the HTR-10 based on its Operation History

    International Nuclear Information System (INIS)

    Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

    2014-01-01

    Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

  15. The challenge of introducing HTR plants on to the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    The international power plant market today is characterized by high increase in energy consumption for developing countries with limitations of investment capital and low increase in energy consumption for industrialized countries with limitations of additional power plant capacities. As a consequence there is a low demand for large new power stations. This leads to a tendency for small and medium sized power plant units - meeting high environmental standards - for which the total investment volume is low and full load operation of a plant can be realized earlier due to the small block capacity. - For nuclear power plants the High-Temperature-Reactor (HTR)-line with spherical fuel elements and a core structure of graphite is specially suited for this small and medium sized nuclear reactor (SMSNR) capacity. The excellent safety characteristics, the high availability, the low radiation doses for the operation personnel and the environment of the HTR line has been demonstrated by 20 years of operation of the AVR-15 MWe experimental power plant in Juelich F.R.G. and since 1985 by operation of the THTR-300 MWe prototype plant at Hamm-Uentrop F.R.G. Up-dated concepts of the HTR-line are under design for electricity generation (HTR-500), for co-generation of power and heat (HTR-100) and for district heating purposes only (GHR-10). By implementing two HTR projects the Brown Boveri Group is in the position to realize the collected experiences from design, licensing, erection, commissioning and operation for the follow-on projects. This leads to practical and sound technical solutions convenient for existing manufacturing processes, well known materials, standardized components and usual manufacturing tolerances. Specific plant characteristics can be used for advantages in the competition. (author)

  16. Reactivity control in HTR power plants with respect to passive safety system. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

  17. Influence of Polymorphisms in the HTR3A and HTR3B Genes on Experimental Pain and the Effect of the 5-HT3 Antagonist Granisetron.

    Science.gov (United States)

    Louca Jounger, Sofia; Christidis, Nikolaos; Hedenberg-Magnusson, Britt; List, Thomas; Svensson, Peter; Schalling, Martin; Ernberg, Malin

    2016-01-01

    The aim of this study was to investigate experimentally if 5-HT3 single nucleotide polymorphisms (SNP) contribute to pain perception and efficacy of the 5-HT3-antagonist granisetron and sex differences. Sixty healthy participants were genotyped regarding HTR3A (rs1062613) and HTR3B (rs1176744). First, pain was induced by bilateral hypertonic saline injections (HS, 5.5%, 0.2 mL) into the masseter muscles. Thirty min later the masseter muscle on one side was pretreated with 0.5 mL granisetron (1 mg/mL) and on the other side with 0.5 mL placebo (isotonic saline) followed by another HS injection (0.2 mL). Pain intensity, pain duration, pain area and pressure pain thresholds (PPTs) were assessed after each injection. HS evoked moderate pain, with higher intensity in the women (P = 0.023), but had no effect on PPTs. None of the SNPs influenced any pain variable in general, but compared to men, the pain area was larger in women carrying the C/C (HTR3A) (P = 0.015) and pain intensity higher in women with the A/C alleles (HTR3B) (P = 0.019). Pre-treatment with granisetron reduced pain intensity, duration and area to a lesser degree in women (P granisetron. Women carrying the C/T & T/T (HTR3A) genotype had less reduction of pain intensity (P = 0.041) and area (P = 0.005), and women with the C/C genotype (HTR3B) had less reduction of pain intensity (P = 0.030), duration (P = 0.030) and area compared to men (P = 0.017). In conclusion, SNPs did not influence experimental muscle pain or the effect of granisetron on pain variables in general, but there were some sex differences in pain variables that seem to be influenced by genotypes. However, due to the small sample size further research is needed before any firm conclusions can be drawn.

  18. Influence of Polymorphisms in the HTR3A and HTR3B Genes on Experimental Pain and the Effect of the 5-HT3 Antagonist Granisetron.

    Directory of Open Access Journals (Sweden)

    Sofia Louca Jounger

    Full Text Available The aim of this study was to investigate experimentally if 5-HT3 single nucleotide polymorphisms (SNP contribute to pain perception and efficacy of the 5-HT3-antagonist granisetron and sex differences. Sixty healthy participants were genotyped regarding HTR3A (rs1062613 and HTR3B (rs1176744. First, pain was induced by bilateral hypertonic saline injections (HS, 5.5%, 0.2 mL into the masseter muscles. Thirty min later the masseter muscle on one side was pretreated with 0.5 mL granisetron (1 mg/mL and on the other side with 0.5 mL placebo (isotonic saline followed by another HS injection (0.2 mL. Pain intensity, pain duration, pain area and pressure pain thresholds (PPTs were assessed after each injection. HS evoked moderate pain, with higher intensity in the women (P = 0.023, but had no effect on PPTs. None of the SNPs influenced any pain variable in general, but compared to men, the pain area was larger in women carrying the C/C (HTR3A (P = 0.015 and pain intensity higher in women with the A/C alleles (HTR3B (P = 0.019. Pre-treatment with granisetron reduced pain intensity, duration and area to a lesser degree in women (P < 0.05, but the SNPs did not in general influence the efficacy of granisetron. Women carrying the C/T & T/T (HTR3A genotype had less reduction of pain intensity (P = 0.041 and area (P = 0.005, and women with the C/C genotype (HTR3B had less reduction of pain intensity (P = 0.030, duration (P = 0.030 and area compared to men (P = 0.017. In conclusion, SNPs did not influence experimental muscle pain or the effect of granisetron on pain variables in general, but there were some sex differences in pain variables that seem to be influenced by genotypes. However, due to the small sample size further research is needed before any firm conclusions can be drawn.

  19. Reprocessing of spent plasma

    International Nuclear Information System (INIS)

    Pierini, G.

    1981-01-01

    This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

  20. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  1. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Nickel, H.

    1986-12-01

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.) [de

  2. European energy policy and the potential impact of HTR and nuclear cogeneration

    International Nuclear Information System (INIS)

    Fütterer, Michael A.; Carlsson, Johan; Groot, Sander de; Deffrennes, Marc; Bredimas, Alexandre

    2014-01-01

    This paper first provides an update on the current state of play and the potential future role of nuclear energy in Europe. It then describes the EU energy policy tools in the area of nuclear technology. It explains the three-tier strategy of the European nuclear technology platform and its demonstration initiatives, here specifically for nuclear cogeneration and HTR. The paper closes with an outlook on the boundary conditions at which HTR can become attractive for nuclear cogeneration, not only from an energy policy viewpoint but also economically

  3. Burner and dissolver off-gas treatment in HTR fuel reprocessing

    International Nuclear Information System (INIS)

    Barnert-Wiemer, H.; Heidendael, M.; Kirchner, H.; Merz, E.; Schroeder, G.; Vygen, H.

    1979-01-01

    In the reprocessing of HTR fuel, essentially all of the gaseous fission products are released during the heat-end tratment, which includes burning of the graphite matrix and dissolving of the heavy metallic residues in THOREX reagent. Three facilities for off-gas cleaning are described, the status of the facility development and test results are reported. Hot tests with a continuous dissolver for HTR-type fuel (throughput 2 kg HM/d) with a closed helium purge loop have been carried out. Preliminary results of these experiments are reported

  4. The conceptual flowsheet of effluent treatment during preparing spherical fuel elements of HTR

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Quan, E-mail: quanying@tsinghua.edu.cn; Xiao-tong, Chen; Bing, Liu; Gen-na, Fu; Yang, Wang; You-lin, Shao; Zhen-ming, Lu; Ya-ping, Tang; Chun-he, Tang

    2014-05-01

    High temperature gas-cooled reactor (HTR) is one of the advanced nuclear reactors owing to its inherent safety and broad applications. For HTR, one of the key components is the ceramic fuel element. During the preparation of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was established. According to the above treatment process, the uranium concentration was decreased from 200 mg/l to the level of discharged standard.

  5. Review on characterization methods applied to HTR-fuel element components

    International Nuclear Information System (INIS)

    Koizlik, K.

    1976-02-01

    One of the difficulties which on the whole are of no special scientific interest, but which bear a lot of technical problems for the development and production of HTR fuel elements is the proper characterization of the element and its components. Consequently a lot of work has been done during the past years to develop characterization procedures for the fuel, the fuel kernel, the pyrocarbon for the coatings, the matrix and graphite and their components binder and filler. This paper tries to give a status report on characterization procedures which are applied to HTR fuel in KFA and cooperating institutions. (orig.) [de

  6. The properties of spherical fuel elements and its behavior in the modular HTR

    International Nuclear Information System (INIS)

    Lohnert, G.H.; Ragoss, H.

    1985-01-01

    The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

  7. Fission Product Releases from a Core into a Coolant of a Prismatic 350-MWth HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A prismatic 350-MW{sub th} high temperature reactor (HTR) is a means to generate electricity and process heat for hydrogen production. The HTR will be operated for an extended fuel burnup of more than 150 GWd/MTU. Korea Atomic Energy Research Institute (KAERI) is performing a point design for the HTR which is a pre-conceptual design for the analysis and assessment of engineering feasibility of the reactor. In a prismatic HTR, metallic and gaseous fission products (FPs) are produced in the fuel, moved through fuel materials, and released into a primary coolant. The FPs released into the coolant are deposited on the various helium-wetted surfaces in the primary circuit, or they are sorbed on particulate matters in the primary coolant. The deposited or sorbed FPs are released into the environment through the leakage or venting of the primary coolant. It is necessary to rigorously estimate such radioactivity releases into the environment for securing the health and safety of the occupational personnel and the public. This study treats the FP releases from a core into a coolant of a prismatic 350-MW{sub th} HTR. These results can be utilized as input data for the estimation of FP migration from a coolant into the environment. The analysis of fission product release within a prismatic 350-MW{sub th} HTR has been done. It was assumed that the HTR was operated at constant temperature and power for 1500 EFPDs. - The final burnup is 152 GWd/tHM at packing fraction of 25 %, and the final fast fluence is about 8 X 10{sup 21} n/cm{sup 2}, E{sub n} > 0.1 MeV. - The temperatures at the compact center and at the center of a kernel located at the compact center are 884 and 893 .deg. C, respectively, when the packing fraction is 25 % and the coolant temperature is 850 .deg. C. - Xenon is the most radioactive fission product in a coolant of a prismatic HTR when there are broken TRISOs and fuel component contaminated with heavy metals. For metallic fission products, the radioactivity

  8. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

  9. European energy policy and the potential impact of HTR and nuclear cogeneration

    Energy Technology Data Exchange (ETDEWEB)

    Fütterer, Michael A., E-mail: michael.fuetterer@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755ZG Petten (Netherlands); Carlsson, Johan [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755ZG Petten (Netherlands); Groot, Sander de [Nuclear Research and consultancy Group, NL-1755ZG Petten (Netherlands); Deffrennes, Marc [European Commission, DG ENER, L-2530 Luxembourg (Luxembourg); Bredimas, Alexandre [LGI Consulting, 13 rue Marivaux, F-75002 Paris (France)

    2014-05-01

    This paper first provides an update on the current state of play and the potential future role of nuclear energy in Europe. It then describes the EU energy policy tools in the area of nuclear technology. It explains the three-tier strategy of the European nuclear technology platform and its demonstration initiatives, here specifically for nuclear cogeneration and HTR. The paper closes with an outlook on the boundary conditions at which HTR can become attractive for nuclear cogeneration, not only from an energy policy viewpoint but also economically.

  10. Chlamydia trachomatis responds to heat shock, penicillin induced persistence, and IFN-gamma persistence by altering levels of the extracytoplasmic stress response protease HtrA

    Directory of Open Access Journals (Sweden)

    Mathews Sarah A

    2008-11-01

    Full Text Available Abstract Background Chlamydia trachomatis, an obligate intracellular human pathogen, is the most prevalent bacterial sexually transmitted infection worldwide and a leading cause of preventable blindness. HtrA is a virulence and stress response periplasmic serine protease and molecular chaperone found in many bacteria. Recombinant purified C. trachomatis HtrA has been previously shown to have both activities. This investigation examined the physiological role of Chlamydia trachomatis HtrA. Results The Chlamydia trachomatis htrA gene complemented the lethal high temperature phenotype of Escherichia coli htrA- (>42°C. HtrA levels were detected to increase by western blot and immunofluorescence during Chlamydia heat shock experiments. Confocal laser scanning microscopy revealed a likely periplasmic localisation of HtrA. During penicillin induced persistence of Chlamydia trachomatis, HtrA levels (as a ratio of LPS were initially less than control acute cultures (20 h post infection but increased to more than acute cultures at 44 h post infection. This was unlike IFN-γ persistence where lower levels of HtrA were observed, suggesting Chlamydia trachomatis IFN-γ persistence does not involve a broad stress response. Conclusion The heterologous heat shock protection for Escherichia coli, and increased HtrA during cell wall disruption via penicillin and heat shock, indicates an important role for HtrA during high protein stress conditions for Chlamydia trachomatis.

  11. Establishment of quality control technology for HTR fuel in Korea

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Kim, Woong Ki; Kim, Yeon Ku; Cho, Moon Sung

    2009-01-01

    Korea is currently developing the HTR coated particle fuel technology in view of its long-term Nuclear Hydrogen Production Technology Development and Demonstration (NHDD) Project, which was launched in 2004, of an extensive R and D program on technology development for a hydrogen production by a VHTR. The current NHDD Project essentially covers the R and D works on the core and reactor system analysis, thermo-hydraulics and safety, coated particle fuel technology, material and component aspects and the hydrogen production technology by using the so-called Sulfur-Iodine Process (S-I Process). As a part of the NHDD Project, the fundamental technology for the coated particle fuel has been being developed, which consist of UO 2 kernel fabrication, pyrolytic carbon (PyC) and silicon carbide (SiC) coating technology, an in-reactor performance model development of a coated particle fuel and a preliminary preparative study for the irradiation tests of the coated particle fuel specimens in the HANARO reactor. In parallel with the development of fabrication process technology of the coated particle fuel, namely, kernel fabrication and coating processes, the characterization techniques for the important characteristics and quality control (QC) methods of the products after each process step were established. This paper deals with the works carried out for the development of the characterization technologies and establishment of the QC techniques for the coated fuel particles. Emphasis is given to the selection and development of the laboratory equipment and apparatus for the development of the methods of the characterizations and relevant QC methods

  12. Increased expression of Apo-J and Omi/HtrA2 after Intracerebral Hemorrage in rats.

    Science.gov (United States)

    Li, Feng; Yang, Jing; Guo, Xiaoyan; Zheng, Xiaomei; Lv, Zhiyu; Shi, Chang Qing; Li, Xiaogang

    2018-03-23

    To investigate the changes of Apo-J and Omi/HtrA2 protein expression in rats with intracerebral hemorrage. 150 SD adult rats were randomly divided into 3 groups: (1) Normal Control (NC) group, (2) Sham group, (3) Intracerebral Hemorrage (ICH) group. The data were collected at 6h, 12h, 1d, 2d, 3d, 5d and 7d. Apoptosis was measured by Tunel staining. The distributions of the Apo-J and Omi/HtrA2 proteins were determined by immunohistochemical staining. The levels of Apo-J mRNA and Omi/HtrA2 mRNA expressions were examined by RT-PCR. Apoptosis in ICH group was higher than Sham and NC groups (p<0.05). Both the Apo-J and Omi/HtrA2 expression levels were increased in the peripheral region of hemorrhage, with a peak at 3d. The Apo-J mRNA level positively correlated with HtrA2 mRNA level in ICH group (r=0.883, p<0.001). The expressions of Apo-J and Omi/HtrA2 paralelly increased in peripheral region of rat cerebral hemorrhage. Local high expressed Apo-J in the peripheral regions might play a neuroprotective role by inhibiting apoptosis via Omi/HtrA2 pathway after hemorrhage. Copyright © 2018. Published by Elsevier Inc.

  13. Identification of E-cadherin signature motifs functioning as cleavage sites for Helicobacter pylori HtrA

    Science.gov (United States)

    Schmidt, Thomas P.; Perna, Anna M.; Fugmann, Tim; Böhm, Manja; Jan Hiss; Haller, Sarah; Götz, Camilla; Tegtmeyer, Nicole; Hoy, Benjamin; Rau, Tilman T.; Neri, Dario; Backert, Steffen; Schneider, Gisbert; Wessler, Silja

    2016-03-01

    The cell adhesion protein and tumour suppressor E-cadherin exhibits important functions in the prevention of gastric cancer. As a class-I carcinogen, Helicobacter pylori (H. pylori) has developed a unique strategy to interfere with E-cadherin functions. In previous studies, we have demonstrated that H. pylori secretes the protease high temperature requirement A (HtrA) which cleaves off the E-cadherin ectodomain (NTF) on epithelial cells. This opens cell-to-cell junctions, allowing bacterial transmigration across the polarised epithelium. Here, we investigated the molecular mechanism of the HtrA-E-cadherin interaction and identified E-cadherin cleavage sites for HtrA. Mass-spectrometry-based proteomics and Edman degradation revealed three signature motifs containing the [VITA]-[VITA]-x-x-D-[DN] sequence pattern, which were preferentially cleaved by HtrA. Based on these sites, we developed a substrate-derived peptide inhibitor that selectively bound and inhibited HtrA, thereby blocking transmigration of H. pylori. The discovery of HtrA-targeted signature sites might further explain why we detected a stable 90 kDa NTF fragment during H. pylori infection, but also additional E-cadherin fragments ranging from 105 kDa to 48 kDa in in vitro cleavage experiments. In conclusion, HtrA targets E-cadherin signature sites that are accessible in in vitro reactions, but might be partially masked on epithelial cells through functional homophilic E-cadherin interactions.

  14. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  15. Spent fuel dissolution mechanisms

    International Nuclear Information System (INIS)

    Ollila, K.

    1993-11-01

    This study is a literature survey on the dissolution mechanisms of spent fuel under disposal conditions. First, the effects of radiolysis products on the oxidative dissolution mechanisms and rates of UO 2 are discussed. These effects have mainly been investigated by using electrochemical methods. Then the release mechanisms of soluble radionuclides and the dissolution of the UO 2 matrix including the actinides, are treated. Experimental methods have been developed for measuring the grain-boundary inventories of radionuclides. The behaviour of cesium, strontium and technetium in leaching tests shows different trends. Comparison of spent fuel leaching data strongly suggests that the release of 90 Sr into the leachant can be used as a measure of the oxidation/dissolution of the fuel matrix. Approaches to the modelling UO 2 , dissolution are briefly discussed in the next chapter. Lastly, the use of natural material, uraninite, in the evaluation of the long-term performance of spent fuel is discussed. (orig.). (81 ref., 37 figs., 8 tabs.)

  16. Shoulder replacement - discharge

    Science.gov (United States)

    Total shoulder arthroplasty - discharge; Endoprosthetic shoulder replacement - discharge; Partial shoulder replacement - discharge; Partial shoulder arthroplasty - discharge; Replacement - shoulder - discharge; Arthroplasty - shoulder - discharge

  17. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1984-02-01

    This paper describes methods and measurements developed at the Austrian Research Centre Seibersdorf for the evaluation of the irradiation performance of HTR fuel. Main interest is concentrated on particle failure rates, fission product release, burn-up and inventory measurements (solid and gaseous fission products, uranium inventory). (Author) [de

  18. Studies on equilibrium fuel management schemes on the Dragon HTR core design

    Energy Technology Data Exchange (ETDEWEB)

    Daub, J; Pedersen, J

    1971-02-03

    The Dragon Project has recently started investigations on fuel management in HTR's with the assumed Dragon design. The study covers the results of investigations into a number of equilibrium fuel management schemes with the 1-dimensional FLATTER code and calculations of the corresponding total power generating costs with the programme TECO.

  19. Standard and chances of the HTR in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Simon, M.; Harder, H.

    1980-01-01

    If one believes the verbal support of the politicians on the High-Temperature Reactor this reactor type seems to have a good future. The facts show that inspite of the well known properties of the HTR and the engagement of the industry and potential operators, progress is hardly made. Soon it could be too late for this reactor line. (orig.) [de

  20. Procedures and results of the probabilistic safety study of the HTR-1160 plant

    International Nuclear Information System (INIS)

    Kroeger, W.; Bongartz, R.

    1981-01-01

    A research team of the Institute for Nuclear Safety Research of the Juelich Nuclear Research Center (KFA) and staff members of the Gesellschaft fuer Reaktorsicherheit, sponsored by the Federal Ministry of the Interior, carried out a safety and risk analysis of high temperature reactors. The studies, which included the transfer to German conditions and the modification in some points of methodology of the American AIPA Study, were performed on the German concept of an 1160 MWe HTR with block-type fuel elements. They referred to accidents and possible impacts on the environment, residual risks and measures to reduce them. The study covered a total of approx. 15 groups of initiating events, including potential external impacts. The dominating initiating events are transients in a pressurized reactor. Differences relative to the light water reactor concept arise with respect to releases as a result of accidents and, above all, release times; they are due to different physical characteristics. HTR'S are characterized by thermal inertia and resistance to temperatures. If the results of the study are extended to the HTR line with a pebble bed core currently in the planning phase, the power densities alone, which are considerably lower in some designs, are indicative of an even more effective fission product retention than is already found in the HTR-1160 plant analyzed here. (orig.) [de

  1. Oxidation of carbon based material for innovative energy systems (HTR, fusion reactor): status and further needs

    International Nuclear Information System (INIS)

    Moormann, R.; Hinssen, H.K.; Latge, Ch.; Dumesnil, J.; Veltkamp, A.C.; Grabon, V.; Beech, D.; Buckthorpe, D.; Dominguez, T.; Krussenberg, A.K.; Wu, C.H.

    2000-01-01

    Following an overview on kinetics of carbon/gas reactions, status and further needs in selected safety relevant fields of graphite oxidation in high temperature reactors (HTRs) and fusion reactors are outlined. Kinetics was detected due to the presence of such elements as severe air ingress, lack of experimental data on Boudouard reaction and a similar lack of data in the field of advanced oxidation. The development of coatings which protect against oxidation should focus on stability under neutron irradiation and on the general feasibility of coatings on HTR pebble fuel graphite. Oxidation under normal operation of direct cycle HTR requires examinations of gas atmospheres and of catalytic effects. Advanced carbon materials like CFCs and mixed materials should be developed and tested with respect to their oxidation resistance in a common HTR/fusion task. In an interim HTR, fuel storage radiolytic oxidation under normal operation and thermal oxidation in accidents have to be considered. Plans for future work in these fields are described. (authors)

  2. A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

    Science.gov (United States)

    Syarip; Po, L. C. C.

    2018-05-01

    In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

  3. Intercomparison of rod-worth measurement techniques in a LEU-HTR assembly

    International Nuclear Information System (INIS)

    Williams, T.; Chawla, R.

    1994-01-01

    The measurement of absorber-rod worths in the radial reflector of a LEU-HTR pebble bed system is described. Particular emphasis is placed on the choice of complementary measurement techniques to ensure that sensitivities to systematic errors in the calculated parameters used in the analysis are minimised. (author) 3 figs., 3 tabs., 8 refs

  4. Viability of HTR-10 as a Primary Driver of an Energy Complex for Remote Settlement

    International Nuclear Information System (INIS)

    Choong, Philip T.

    2014-01-01

    HTR-10, a proven 10 MWt prototype pebble bed reactor, is capable of generating 4 MWe to the power grid. However; with evolutional power upgrades, its output performance can be substantially enhanced to drive an energy complex to co-generate electricity, hydrogen, desalinated water and process heat for a remote island or settlement of several thousand people. Unlike the much publicized SMR power concepts in the literature, HTR-10 is the only full-blown stand-alone power system that has been demonstrated to be inherently safe and capable of high temperature output. Furthermore, this particular HTR family of reactors is proliferation-resistant and possesses many desirable market-competitive advantages such as high thermal efficiency, low thermal pollution, zero carbon footprints and minimal exclusion zones. An innovative classroom project course is structured to stimulate science and engineering students to explore novel use of HTR-10 as a high temperature heat source to be the core of an intelligent zero emission energy (Smart-ZEE) module capable of providing all energy needs of a remote community or island. (author)

  5. Reracking to increase spent fuel storage capacity

    International Nuclear Information System (INIS)

    1980-05-01

    Many utilities have already increased their spent fuel pool storage capacity by replacing aluminum racks having storage densities as low as 0.2 MTU/ft 2 with stainless steel racks which can more than double storage densities. Use of boron-stainless steel racks or thin stainless steel cans containing reassembled fuel rods allows even higher fuel storage densities (up to approximately 1.25 MTU/ft 2 ). This report evaluates the economics of smaller storage gains that occur if pools, already converted to high density storage, are further reracked

  6. Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

    Science.gov (United States)

    Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

  7. Dual regulatory switch confers tighter control on HtrA2 proteolytic activity.

    Science.gov (United States)

    Singh, Nitu; D'Souza, Areetha; Cholleti, Anuradha; Sastry, G Madhavi; Bose, Kakoli

    2014-05-01

    High-temperature requirement protease A2 (HtrA2), a multitasking serine protease that is involved in critical biological functions and pathogenicity, such as apoptosis and cancer, is a potent therapeutic target. It is established that the C-terminal post-synaptic density protein, Drosophila disc large tumor suppressor, zonula occludens-1 protein (PDZ) domain of HtrA2 plays pivotal role in allosteric modulation, substrate binding and activation, as commonly reported in other members of this family. Interestingly, HtrA2 exhibits an additional level of functional modulation through its unique N-terminus, as is evident from 'inhibitor of apoptosis proteins' binding and cleavage. This phenomenon emphasizes multiple activation mechanisms, which so far remain elusive. Using conformational dynamics, binding kinetics and enzymology studies, we addressed this complex behavior with respect to defining its global mode of regulation and activity. Our findings distinctly demonstrate a novel N-terminal ligand-mediated triggering of an allosteric switch essential for transforming HtrA2 to a proteolytically competent state in a PDZ-independent yet synergistic activation process. Dynamic analyses suggested that it occurs through a series of coordinated structural reorganizations at distal regulatory loops (L3, LD, L1), leading to a population shift towards the relaxed conformer. This precise synergistic coordination among different domains might be physiologically relevant to enable tighter control upon HtrA2 activation for fostering its diverse cellular functions. Understanding this complex rheostatic dual switch mechanism offers an opportunity for targeting various disease conditions with tailored site-specific effector molecules. © 2014 FEBS.

  8. Plant Operation Station for HTR-PM Low Power and Shutdown operation Probabilistic safety analysis

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan

    2014-01-01

    Full range Probabilistic safety analysis (PSA) is one of key conditions for nuclear power plant (NPP) licensing according to the requirement of nuclear safety regulatory authority. High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) has developed construction design and prepared for the charging license application. So after the normal power operation PSA submitted for review, the Low power and Shutdown operation Probabilistic safety analysis (LSPSA) also begin. The results of LSPSA will together with prior normal power PSA results to demonstrate the safety level of HTR-PM NPP Plant Operation Station (POS) is one of important terms in LSPSA. The definition of POS lays the foundation for LSPSA modeling. POS provides initial and boundary conditions for the following event tree and fault tree model development. The aim of this paper is to describe the state-of-the-art of POS definition for HTR-PM LSPSA. As for the first attempt to the high temperature gas cooled reactor module plant, the methodology and procedure of POS definition refers to the LWR LSPSA guidance, and adds to plant initial status analysis due to the HTR-PM characteristics. A specific set of POS grouping vectors is investigate and suggested for HTR-PM NPP, which reflects the characteristics of plant modularization and on-line refueling. As a result, seven POSs are given according to the grouping vectors at the end of the paper. They will be used to the LSPSA modelling and adjusted if necessary. The papers ’work may provide reference to the analogous NPP LSPSA. (author)

  9. MTR radiological database for SRS spent nuclear fuel facilities

    International Nuclear Information System (INIS)

    Blanchard, A.

    2000-01-01

    A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

  10. Pinhole Breaches in Spent Fuel Containers: Some Modeling Considerations

    International Nuclear Information System (INIS)

    Casella, Andrew M.; Loyalka, Sudarsham K.; Hanson, Brady D.

    2006-01-01

    This paper replaces PNNL-SA-48024 and incorporates the ANS reviewer's comments, including the change in the title. Numerical methods to solve the equations for gas diffusion through very small breaches in spent fuel containers are presented and compared with previous literature results

  11. Hip joint replacement

    Science.gov (United States)

    Hip arthroplasty; Total hip replacement; Hip hemiarthroplasty; Arthritis - hip replacement; Osteoarthritis - hip replacement ... Your hip joint is made up of 2 major parts. One or both parts may be replaced during surgery: ...

  12. Spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.

    1982-06-01

    Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR, PWR) within a given utility. In all cases, a full core discharge capability (full core reserve or FCR) is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. For the current AR capacity case the indicated storage requirements in the year 2000 are indicated to be 18,190 MTU; for the maximum capacity with transshipment case they are 11,320 MTU

  13. Spent fuel transportation problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.A.

    1977-01-01

    In this paper, problems of transportation of nuclear spent fuel to reprocessing plants are discussed. The solutions proposed are directed toward the achievement of the transportation as economic and safe as possible. The increase of the nuclear power plants number in the USSR and the great distances between these plants and the reprocessing plants involve an intensification of the spent fuel transportation. Higher burnup and holdup time reduction cause the necessity of more bulky casks. In this connection, the economic problems become still more important. One of the ways of the problem solution is the development of rational and cheap cask designs. Also, the enforcement in the world of the environmental and personnel health protection requires to increase the transportation reliability and safety. The paper summarizes safe transportation rules with clarifying the following questions: the increase of the transport unit quantity of the spent fuel; rational shipment organization that minimizes vehicle turnover cycle duration; development of the reliable calculation methods to determine strength, thermal conditions and nuclear safety of transport packaging as applied to the vehicles of high capacity; maximum unification of vehicles, calculation methods and documents; and cask testing on models and in pilot scale on specific test rigs to assure that they meet the international safe fuel shipment rules. Besides, some considerations on the choice and use of structural materials for casks are given, and problems of manufacturing such casks from uranium and lead are considered, as well as problems of the development of fireproof shells, control instrumentation, vehicles decontamination, etc. All the problems are considered from the point of view of normal and accidental shipment conditions. Conclusions are presented [ru

  14. Numerical calculation and analysis of natural convection removal of the spent fuel residual heat of 10 MW high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Wang Jinhua; Huang Yifan; Wu Bin

    2013-01-01

    The spent fuel of 10 MW High Temperature Gas Cooled Reactor (HTR-10) could be stored in the shielded tank, and the tank is stored in the concrete shielded canister in spent fuel storage room, the residual heat of the spent fuel could be removed by the air. The ability of residual heat removal is analyzed in the paper, and the temperature field is numerically calculated through FEA program ANSYS, the analysis and the calculation are used to validate the safety of the spent fuel and the tank, the ultimate temperature of the spent fuel and the tank should below the safety limit. The calculation shows that the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank, and the temperature decreases gradually with radial distance, the temperature in the tank body is evenly distributed, and the temperature in the concrete shielded canister decreases gradually with radial distance. It is feasible to remove the residual heat of the spent fuel storage tank by natural ventilation, in natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method. (authors)

  15. Distinct 3D Architecture and Dynamics of the Human HtrA2(Omi Protease and Its Mutated Variants.

    Directory of Open Access Journals (Sweden)

    Artur Gieldon

    Full Text Available HtrA2(Omi protease controls protein quality in mitochondria and plays a major role in apoptosis. Its HtrA2S306A mutant (with the catalytic serine routinely disabled for an X-ray study to avoid self-degradation is a homotrimer whose subunits contain the serine protease domain (PD and the regulatory PDZ domain. In the inactive state, a tight interdomain interface limits penetration of both PDZ-activating ligands and PD substrates into their respective target sites. We successfully crystalized HtrA2V226K/S306A, whose active counterpart HtrA2V226K has had higher proteolytic activity, suggesting higher propensity to opening the PD-PDZ interface than that of the wild type HtrA2. Yet, the crystal structure revealed the HtrA2V226K/S306A architecture typical of the inactive protein. To get a consistent interpretation of crystallographic data in the light of kinetic results, we employed molecular dynamics (MD. V325D inactivating mutant was used as a reference. Our simulations demonstrated that upon binding of a specific peptide ligand NH2-GWTMFWV-COOH, the PDZ domains open more dynamically in the wild type protease compared to the V226K mutant, whereas the movement is not observed in the V325D mutant. The movement relies on a PDZ vs. PD rotation which opens the PD-PDZ interface in a lid-like (budding flower-like in trimer fashion. The noncovalent hinges A and B are provided by two clusters of interfacing residues, harboring V325D and V226K in the C- and N-terminal PD barrels, respectively. The opening of the subunit interfaces progresses in a sequential manner during the 50 ns MD simulation. In the systems without the ligand only minor PDZ shifts relative to PD are observed, but the interface does not open. Further activation-associated events, e.g. PDZ-L3 positional swap seen in any active HtrA protein (vs. HtrA2, were not observed. In summary, this study provides hints on the mechanism of activation of wtHtrA2, the dynamics of the inactive HtrA2V325D

  16. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  17. Spent fuel interim storage

    International Nuclear Information System (INIS)

    Bilegan, Iosif C.

    2003-01-01

    The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

  18. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Kurokawa, Hideaki; Kumagaya, Naomi; Oda, Masashi; Matsuda, Masami; Maruyama, Hiromi; Yamanaka, Tsuneyasu.

    1997-01-01

    The structure of a spent fuel storage rack is determined by the material, thickness, size of square cylindrical tubes (the gap between spent fuel assemblies and the square cylindrical tubes) and pitch of the arrangement (the gap between each of the square cylindrical tubes). In the present invention, the thickness and the pitch of the arrangement of the square tubes are optimized while evaluating subcriticality. Namely, when the sum of the thickness of the water gap at the outer side (the pitch of arrangement of the cylindrical tubes) and the thickness of the cylindrical tubes is made constant, the storage rack is formed by determining the thickness of the cylindrical tubes which is smaller than the optimum value among the combination of the thickness of the water gap at the outer side and that of the cylindrical tube under the effective multiplication factor to be performed. Then, the weight of the rack can be reduced, and the burden of the load on the bottom of the pool can be reduced. Further, the amount of the constitutional materials of the rack itself can be reduced thereby capable of reducing the cost for the materials of the rack. (T.M.)

  19. Development status of the HTGR in the world. Outline and construction status of the demonstration HTGR program (HTR-PM) of China

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Okamoto, Futoshi; Mouri, Tomoaki; Saito, Masanao; Nishio, Hiroki; Ohashi, Junpei

    2014-01-01

    Based on successful construction and operation experiences of HTR-10 reactor with pebble bed fuel and helium coolant, HTR-PM (HTR Pebble-bed Modular) reactor program was under way with 200 MWe of twin reactors with the same core configuration as HTR-10 reactor, which, each with a single steam generator, would drive a single steam turbine. Core height was 11 meters, and main steam temperature would be at 566 C. Although HTR-PM reactor program was interrupted by effects of the Fukushima accident, first concrete basement construction was started in December 2012 with aiming at connecting the Grid in 2017. This article reviewed outline and construction status of HTR-PM reactor in China. (T. Tanaka)

  20. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  1. Spent fuel: prediction model development

    International Nuclear Information System (INIS)

    Almassy, M.Y.; Bosi, D.M.; Cantley, D.A.

    1979-07-01

    The need for spent fuel disposal performance modeling stems from a requirement to assess the risks involved with deep geologic disposal of spent fuel, and to support licensing and public acceptance of spent fuel repositories. Through the balanced program of analysis, diagnostic testing, and disposal demonstration tests, highlighted in this presentation, the goal of defining risks and of quantifying fuel performance during long-term disposal can be attained

  2. Conception of a modular HTR-process heat facility with optimization of the pressure level

    International Nuclear Information System (INIS)

    Bousack, H.

    1984-11-01

    The operation of a steam reformer heated by nuclear power with a process pressure of about 20 bar provides advantages with respect to process engineering due to the improved conversion and simplified product gas treatment for the follow-on process. The effects of a reduction in pressure on the components of the primary circuit in a modular HTR facility, as well as various process engineering possibilities for producing methanol in the follow-on process are discussed in this paper. Studies cover the influence of core geometry and power density, as well as possibilities of increasing the modular power at a maximum accident temperature of 1600 0 C. An inherently functioning area cooling system is proposed for afterheat removal outside the primary circuit. Based on the optimized pressure, a modular HTR process heat facility is conceived to produce methanol from natural gas and carbon dioxide basically satisfying the requirement of zero emission. (orig.) [de

  3. Steroid hormones modulate galectin-1 in the trophoblast HTR-8/SVneocell line

    Directory of Open Access Journals (Sweden)

    Bojić-Trbojević Žanka

    2008-01-01

    Full Text Available The effects of steroids on galectin-1 (gal-1 were studied in HTR-8/SVneo cells by immunocytochemistry, cell-based ELISA, the MTT proliferation test and the Matrigel TM invasion test. Dexamethasone (DEX, progesterone (PRG, and mifepristone (RU486 were used. Gal-1 was modulated in a steroid- and dose-dependent manner by DEX, which mildly but significantly stimulated production at low concentrations (0.1-10 nM, and inhibited it at 100 nM, while the effects of PRG and RU486 were opposite. HTR-8/SVneo cell invasion of Matrigel was significantly decreased in the presence of DEX and lactose. The obtained data support the proposed regulatory role of steroids in trophoblast gal-1 production.

  4. Comparative Study on Electric Generation Cost of HTR with Another Electric Plant Using LEGECOST Program

    International Nuclear Information System (INIS)

    Mochamad-Nasrullah; Soetrisnanto, Arnold Y.; Tosi-Prastiadi; Adiwardojo

    2000-01-01

    Monetary and economic crisis in Indonesia resulted in impact of electricity and demand and supply planning that it has to be reevaluated. One of the reasons is budget limitation of the government as well as private companies. Considering this reason, the economic calculation for all of aspect could be performed, especially the calculation of electric generation cost. This paper will discuss the economic aspect of several power plants using fossil and nuclear fuel including High Temperature Reactor (HTR). Using Levelized Generation Cost (LEGECOST) program developed by IAEA (International Atomic Energy Agency), the electric generation cost of each power plant could be calculated. And then, the sensitivity analysis has to be done using several economic parameters and scenarios, in order to be known the factors that influence the electric generation cost. It could be concluded, that the electric generation cost of HTR is cheapest comparing the other power plants including nuclear conventional. (author)

  5. Development of Chinese HTR-PM pebble bed equivalent conductivity test facility

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Cheng; Yang, Xingtuan; Jiang, Shengyao [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology

    2016-01-15

    The first two 250-MWt high-temperature reactor pebble bed modules (HTR-PM) have been installing at the Shidaowan plant in Shandong Province, China. The values of the effective thermal conductivity of the pebble bed core are essential parameters for the design. For their determination, Tsinghua University in China has proposed a full-scale heat transfer experiment to conduct comprehensive thermal transfer tests in packed pebble bed and to determine the effective thermal conductivity.

  6. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Lako, P.

    1996-06-01

    In this study the economic feasibility of new concepts of the High Temperature Reactor were investigated. These new concepts are characterized as inherently safe. The different concepts were used as industrial heat/power reactors and compared with a gas fired Steam and Gas turbine installation. The best economic advantages are offered by a HTR with a Thorium/Uranium cycle as compared with a gas fired steam- and gas turbine. 6 figs, 9 tabs, 21 refs

  7. An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1982-01-01

    The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

  8. Status of Research on Pebble Bed HTR Fuel Fabrication Technology in Indonesia

    International Nuclear Information System (INIS)

    Rachmawati, M.; Sarjono; Ridwan; Langenati, R.

    2014-01-01

    Research on pebble bed HTR fuel fabrication is conducted in Indonesia. One of the aims is to build a knowledge base on pebble bed HTR fuel element fabrication technology for fuel procurement. The steps of research strategies are firstly to understand the basic design research of TRISO fuel, properties, and requirements, and secondly to understand the TRISO fuel manufacturing technology, which comprises fabrication and quality control, including its facility. Both steps are adopted from research and experiences of the countries with HTR fuel element fabrication technology. From the knowledge gained in the research, an experimental design of the process and a set of prototype process equipment for fabrication are developed, namely kernels production using external gelation process, TRISO coating of the kernel, and pebble compacting. Experiments using the prototypes have been conducted. Characterization of the kernel product, i.e. diameter, sphericity, density and O/U ratio, shows that the kernel product is still not in compliance with the specification requirements. These are deemed to be caused mainly by the selected vibrating system and the viscosity adjustment. Another major cause is the selected NH3 and air feeding method for both NH3 and air layer in the preparation for spherical droplets of liquid. The FB-CVD TRISO coating of the kernel has been experimented but unsuccessful by using an FB-CVD once‐through continuous coating process. For the pebble compacting, the process is still in the early stage of setting-up compaction equipment. This paper summarizes the current status of research on HTR fuel fabrication technology in Indonesia, the proposed process and its equipment setting-up for improvement of the kernel production. The knowledge and lessons learned gained from the research is useful and can be an assistance in planning for fuel development laboratory facilities procurement, formulating User Requirement Document and Bid Invitation Specification for

  9. Analysis of hypothetical incidents in nuclear power plants with PWR and HTR

    International Nuclear Information System (INIS)

    Geiser, H.

    1977-01-01

    Several accident analyses are reviewed with a view to fission product release, and the findings are transferred to German reactor plants with LWR and HTR and compared. First of all, hypothetical accidents are compared for both of these lines; after this, the history of accidents is briefly described, and the fission product release during these accidents is investigated. For both reactor lines, there is a different but sufficiently high potential for safety improvements. (orig.) [de

  10. Study on the shuffling scheme in HTR-10 MW test module

    International Nuclear Information System (INIS)

    Jing Xingqing; Zhang Xu; Luo Jingyu

    1993-01-01

    The shuffling ways, once through then out and multiple through then out, in HTR-10 MW Test Module are studied. Multiple through then out is better than once through with regard to rational use of the fuel and flattening the power. The behaviour of equilibrium core and loss of coolant accident is analyzed. The results indicate that characteristic features of the multiple through then out could be better to satisfy the demands of safety criterions

  11. Transcatheter aortic valve replacement

    Science.gov (United States)

    ... gov/ency/article/007684.htm Transcatheter aortic valve replacement To use the sharing features on this page, please enable JavaScript. Transcatheter aortic valve replacement (TAVR) is surgery to replace the aortic valve. ...

  12. Hip Replacement Surgery

    Science.gov (United States)

    ... Outreach Initiative Breadcrumb Home Health Topics English Español Hip Replacement Surgery Basics In-Depth Download Download EPUB ... PDF What is it? Points To Remember About Hip Replacement Surgery Hip replacement surgery removes damaged or ...

  13. Nicotine replacement therapy

    Science.gov (United States)

    Smoking cessation - nicotine replacement; Tobacco - nicotine replacement therapy ... Before you start using a nicotine replacement product, here are some things to know: The more cigarettes you smoke, the higher the dose you may need to ...

  14. Experimental Study of Fuel Element Motion in HTR-PM Conveying Pipelines

    International Nuclear Information System (INIS)

    Wang Xin; Zhang Haiquan; Nie Junfeng; Li Hongke; Liu Jiguo; He Ayada

    2014-01-01

    The motion action of sphere fuel element (FE) inside fuel pipelines in HTR-PM is indeterminate. Fuel motion is closely connected with the interaction of FE and inner surface of fuel conveying pipe. In this paper, motion method of fuel elements in its conveying pipe is Experimental studied. Combined with the measurement of the fuel passing speed in stainless steel pipe and the track left by sphere ball for experiment, interaction modes of fuel and inner-surface of pipe, which is sliding friction, rolling friction and Collision, has been found. The modes of interaction can affect the speed of fuel conveying, amount of sphere waste and operation stability of fuel handling of high temperature reactor-pebble bed modules (HTR-PM). Furthermore, the motion process of fuel passing a big-elbow which is lying on the top of fuel pneumatic hoisting pipe were experimented. The result shows that the speed before and the speed after the elbow is positive correlation. But with the increase of speed before the elbow, the speed after the elbow increase less. Meanwhile the fuel conveying mode changes from friction to collision. And the conveying process is still steady. The effect can be used to controlling the speed of fuel conveying in fuel handling process of HTR-PM. (author)

  15. Design investigation of the HTR for the opening of very heavy oil deposits

    International Nuclear Information System (INIS)

    Gao, Z.

    1985-02-01

    In the north-east of China there are rich deposits of very heavy oil, which are to be found in a depth of 1500-1700 m. For opening an interaction of 370-390 0 Celsius steam is necessary. The HTR is well suited to produce the steam. A nuclear heat source of 1000 MWsub(th) makes possible the production of 1.5 million tons oil per year. This is a 30-40 per cent higher production of oil compared to the oil-fired steam production. Two concepts of smaller pebble bed reactors are suited as heat sources: the HTR-MEDUL-334 with a thermal power of the 334 MW and fuelled in the multiple run-through scheme and the HTR-OTTO-200 with 200 MW and once-through fuelling. Three or five reactors can be combined in the modular way to provide the power of 1000 MW. For both reactors the design, the neutron-physical and thermohydraulic behaviour are followed in the computer simulation. A central zone of the pebble bed reactor is fuelled with elements of strongly reduced fissile content. Due to the reduced power density the maximum fuel temperature appearing in extreme accidents is limited and accordingly the release of the fission products is avoided. (orig.) [de

  16. Research on Fault Diagnosis of HTR-PM Based on Multilevel Flow Model

    International Nuclear Information System (INIS)

    Zhang Yong; Zhou Yangping

    2014-01-01

    In this paper, we focus on the application of Multilevel Flow Model (MFM) in the automatic real-time fault diagnosis of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) accidents. In the MFM, the plant process is described abstractly in function level by mass, energy and information flows, which reveal the interaction between different components and capacitate the causal reasoning between functions according to the flow properties. Thus, in the abnormal status, a goal-function-component oriented fault diagnosis can be performed with the model at a very quick speed and abnormal alarms can be also precisely explained by the reasoning relationship of the model. By using MFM, a fault diagnosis model of HTR-PM plant is built, and the detailed process of fault diagnosis is also shown by the flowcharts. Due to lack of simulation data about HTR-PM, experiments are not conducted to evaluate the fault diagnosis performance, but analysis of algorithm feasibility and complexity shows that the diagnosis system will have a good ability to detect and diagnosis accidents timely. (author)

  17. The HTR modular power reactor system. Qualification of fuel elements and materials

    International Nuclear Information System (INIS)

    Heidenreich, U.; Breitling, H.; Nieder, R.; Ohly, W.; Mittenkuehler, A.; Ragoss, H.; Seehafer, H.J.; Wirtz, K.; Serafin, N.

    1989-01-01

    For further development of the HTR modular power reactor system (HTR-M-KW), the project activities for 'Qualification of fuel elements and materials' reported here cover the work for specifying the qualifications to be met by metallic and ceramic materials, taking into account the design-based requirements and the engineered safety requirements. The fission product retention data determined for the HTR modular reactor fuel elements could be better confirmed by evaluation of the experiments, and have been verified by various calculation methods for different operating conditions. The qualification of components was verified by strength analyses including a benchmark calculation for specified normal operation and emergencies; the results show a convenient behaviour of the components and their materials. In addition, a fuel element burnup measuring system was designed that applies Cs-137 gamma spectroscopy; its feasibility was checked by appropriate analyses, and qualification work is in progress. The installation of a prototype measurement system is the task for project No. 03 IAT 211. (orig.) [de

  18. Digital Distributed Control System Design: Control Policy for Shared Objects in HTR-PM

    International Nuclear Information System (INIS)

    Zhou Shuqiao; Huang Xiaojin

    2014-01-01

    HTR-PM is an HTR demonstration plant with a structure of two modules feeding one steam turbine. Compared with the structure of one single reactor feeding one turbine, there are more devices shared between these two modules. When they are operated, the shared components are prone to introduce collisions or even logical deadlocks for different technical processes. The future commercial HTR-PM plants are supposed to comprise more modules for a larger turbine, thus the collision problem introduced by the shared components may become severer. Therefore, how to design suitable policies in the distributed control system (DCS) to relieve the collisions during using these shared devices is a new and also a very important problem. In this paper, the classifications of the shared devices are first addressed, and then how to identify the shared objects of an NPP is proposed. Furthermore, a general model for the control logic design is proposed, taking into consideration the collision avoidance, time delay and fairness. The example of how to apply the schemes to relieve the conflicts and deadlocks in the processes of using the shared devices in fuel element cycling system is illustrated. (author)

  19. Possibility of using gamma radiation from HTR reactors for the processing of food and medical products

    International Nuclear Information System (INIS)

    Pahladsingh, R.R.

    2004-01-01

    During the fission process in most of the presently operating nuclear reactors nuclear energy is converted into thermal energy and transferred to common steam cycles for power generation. As part of the fission process also α, β and neutrons particles are released from the nucleus; the release of gamma-rays is also a part of the fission process. In present nuclear reactors α, β, neutrons particles and particularly gamma-rays are not gainfully used as a result of the reactor design and of the containment. These plants are built as required by regulations and international standards for safety. The inherently safe HTR reactor, by its physics and design, does not need a special reinforced containment and it is worth looking into the possibilities of this design feature to use the by-products, such as Gamma-rays, from nuclear fission. In the HTR Pebble Bed Reactors the α, and β particles will remain in the kernels of the pebbles. This means that only the neutron particles and gamma-rays will be available outside the reactor pressure vessel. In this report a proposal is presented to use the gamma-rays of the HTR reactor for irradiation of food and agricultural produce. For neutron shielding a reflector is placed inside the reactor while outside the reactor neutron- and thermal-shielding will be accomplished with water. The high energy gamma-rays will pass through the water-shield and could be harnessed for radiation processing of food and medical products. (author)

  20. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  1. A family-based association study of the HTR1B gene in eating disorders

    Directory of Open Access Journals (Sweden)

    Sandra Hernández

    Full Text Available Objective: To explore the association of three polymorphisms of the serotonin receptor 1Dβ gene (HTR1B in the etiology of eating disorders and their relationship with clinical characteristics. Methods: We analyzed the G861C, A-161T, and A1180G polymorphisms of the HTR1B gene through a family-based association test (FBAT in 245 nuclear families. The sample was stratified into anorexia nervosa (AN spectrum and bulimia nervosa (BN spectrum. In addition, we performed a quantitative FBAT analysis of anxiety severity, depression severity, and Yale-Brown-Cornell Eating Disorders Scale (YBC-EDS in the AN and BN-spectrum groups. Results: FBAT analysis of the A-161T polymorphism found preferential transmission of allele A-161 in the overall sample. This association was stronger when the sample was stratified by spectrums, showing transmission disequilibrium between the A-161 allele and BN spectrum (z = 2.871, p = 0.004. Quantitative trait analysis showed an association between severity of anxiety symptoms and the C861 allele in AN-spectrum participants (z = 2.871, p = 0.004. We found no associations on analysis of depression severity or preoccupation and ritual scores in AN or BN-spectrum participants. Conclusions: Our preliminary findings suggest a role of the HTR1B gene in susceptibility to development of BN subtypes. Furthermore, this gene might have an impact on the severity of anxiety in AN-spectrum patients.

  2. Two-branch Gas Experiments for Hot Gas Mixing of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Yangping; Hao Pengefei; He Heng; Li Fu; Shi Lei

    2014-01-01

    A model experiment is proposed to investigate the hot gas mixing efficiency of HTR-PM reactor outlet. The test facility is introduced which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main body of test facility, hot gas duct, exhaust pipe system and I&C system. Two-branch gas experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis result shows the mixing efficiency is higher than the requirement of thermal mixing by steam generator even with conservative assumption which indicates that the design of hog gas mixing structure of HTR-PM fulfills the requirement for thermal mixing at two-branch working conditions. (author)

  3. Different Roles of COMT and HTR2A Genotypes in Working Memory Subprocesses.

    Directory of Open Access Journals (Sweden)

    Hirohito M Kondo

    Full Text Available Working memory is linked to the functions of the frontal areas, in which neural activity is mediated by dopaminergic and serotonergic tones. However, there is no consensus regarding how the dopaminergic and serotonergic systems influence working memory subprocesses. The present study used an imaging genetics approach to examine the interaction between neurochemical functions and working memory performance. We focused on functional polymorphisms of the catechol-O-methyltransferase (COMT Val(158Met and serotonin 2A receptor (HTR2A -1438G/A genes, and devised a delayed recognition task to isolate the encoding, retention, and retrieval processes for visual information. The COMT genotypes affected recognition accuracy, whereas the HTR2A genotypes were associated with recognition response times. Activations specifically related to working memory were found in the right frontal and parietal areas, such as the middle frontal gyrus (MFG, inferior frontal gyrus (IFG, anterior cingulate cortex (ACC, and inferior parietal lobule (IPL. MFG and ACC/IPL activations were sensitive to differences between the COMT genotypes and between the HTR2A genotypes, respectively. Structural equation modeling demonstrated that stronger connectivity in the ACC-MFG and ACC-IFG networks is related to better task performance. The behavioral and fMRI results suggest that the dopaminergic and serotonergic systems play different roles in the working memory subprocesses and modulate closer cooperation between lateral and medial frontal activations.

  4. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  5. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

    1990-01-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  6. Study on the Break Accidents of the HTR-PM Primary Loop

    International Nuclear Information System (INIS)

    Lang Minggang; Sun Ximing; Zheng Yanhua

    2014-01-01

    In thermal hydraulics design and safety analysis of the HTR-PM, the THERMIX code was used to study the behavior of the helium in the primary system. Once the helium leaks from the primary loop through a break or a relief valve, it is hard to simulate the states of the leakage room with THERMIX. In this paper, the latest version of RELAP5/MOD4, was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 model of the HTR-PM, including the core, the primary system, the secondary loop and the containment, were developed and evaluated in this paper. Based on the model, this paper studied the accidents consequences of a large break in the pressure relief room and a small break in the instrument room of the HTR-PM reactor building. The simulating results illustrate that the temperature in the pressure relief room was no more than 200℃ after a un-isolating large break, and the temperature in the instrument room is less than 130 ℃ after a small un-isolating break. The analysis shows that the scram function and the ability to monitor the reactor temperature and pressure after accidents would not be affected by the break. (author)

  7. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

    1990-07-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  8. Predictions of the Bypass Flows in the HTR-PM Reactor Core

    International Nuclear Information System (INIS)

    Sun Jun; Chen Zhipeng; Zheng Yanhua; Shi Lei; Li Fu

    2014-01-01

    In the HTR-PM reactor core, the basic structure materials are large amount of graphite reflectors and carbon bricks. Small gaps among those graphite and carbon bricks are widespread in the reactor core so that the cold helium flow may be bypassed and not completely heated. The bypass flows in relative lower temperature would change the flow and temperature distributions in the reactor core, therefore, the accurate prediction of bypass flows need to be carried out carefully to evaluate the influence to the reactor safety. Based on the characteristics of the bypass flow problem, hybrid method of the flow network and the CFD tools was employed to represent the connections and calculate flow distributions of all the main flow and bypass flow paths. In this paper, the hybrid method was described and applied to specific bypass flow problem in the HTR-PM. Various bypass flow paths in the HTR-PM were reviewed, figured out, and modeled by the flow network and the CFD methods, including the axial vertical gaps in the side reflectors, control rod channels, absorber sphere channels and radial gap flow through keys around the hot helium plenum. The bypass flow distributions and its flow rate ratio to the total flow rate in the primary loop were also calculated, discussed and evaluated. (author)

  9. Spent Fuel Transfer to Dry Storage Using Unattended Monitoring System

    International Nuclear Information System (INIS)

    Park, Jae Hwan; Park, Soo Jin

    2009-01-01

    There are 4 CANDU reactors at Wolsung site together with a spent fuel dry storage associated with unit 1. These CANDU reactors, classified as On-Load Reactor (OLR) for Safeguards application, change 16- 24 fuel bundles with fresh fuel in everyday. Especially, the spent fuel bundles are transferred from spent fuel bays to dry storage throughout a year because of the insufficient capacity of spent fuel pond. Safeguards inspectors verify the spent fuel transfer to meet safeguards purposes according to the safeguards criteria by means of inspector's presence during the transfer campaign. For the verification, 60-80 person-days of inspection (PDIs) are needed during approximately 3 months for each unit. In order to reduce the inspection effort and operators' burden, an Unattended Monitoring System (UMS) was designed and developed by the IAEA for the verification of spent fuel bundles transfers from wet storage to dry storage. Based on the enhanced cooperation of CANDU reactors between the ROK and the IAEA, the IAEA installed the UMS at Wolsung unit 2 in January 2005 at first. After some field trials during the transfer campaign, this system is being replaced the traditional human inspection since September 1, 2006 combined with a Short Notice Inspection (SNI) and a near-real time Mailbox Declaration

  10. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Mineo, H.; Nomura, Y.; Sakamoto, K.

    1998-01-01

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  11. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

    Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

  12. WWER spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Bower, C C; Lettington, C [GEC Alsthom Engineering Systems Ltd., Whetstone (United Kingdom)

    1994-12-31

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs.

  13. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  14. Spent fuel management overview: a global perspective

    International Nuclear Information System (INIS)

    Bonne, A.; Crijns, M.J.; Dyck, P.H.; Fukuda, K.; Mourogov, V.M.

    1999-01-01

    The paper defines the main spent fuel management strategies and options, highlights the challenges for spent fuel storage and gives an overview of the regional balances of spent fuel storage capacity and spent fuel arising. The relevant IAEA activities in the area of spent fuel management are summarised. (author)

  15. Potential Applications for Nuclear Energy besides Electricity Generation: AREVA Global Perspective of HTR Potential Market

    International Nuclear Information System (INIS)

    Soutworth, Finis; Gauthier, Jean-Claude; Lecomte, Michel; Carre, Franck

    2007-01-01

    Energy supply is increasingly showing up as a major issue for electricity supply, transportation, settlement, and process heat industrial supply including hydrogen production. Nuclear power is part of the solution. For electricity supply, as exemplified in Finland and France, the EPR brings an immediate answer; HTR could bring another solution in some specific cases. For other supply, mostly heat, the HTR brings a solution inaccessible to conventional nuclear power plants for very high or even high temperature. As fossil fuels costs increase and efforts to avoid generation of Greenhouse gases are implemented, a market for nuclear generated process heat will develop. Following active developments in the 80's, HTR have been put on the back burner up to 5 years ago. Light water reactors are widely dominating the nuclear production field today. However, interest in the HTR technology was renewed in the past few years. Several commercial projects are actively promoted, most of them aiming at electricity production. ANTARES is today AREVA's response to the cogeneration market. It distinguishes itself from other concepts with its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important of which is the design flexibility to adapt readily to combined heat and power applications. From the start, AREVA made the choice of such flexibility with the belief that the HTR market is not so much in competition with LWR in the sole electricity market but in the specific added value market of cogeneration and process heat. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source free of greenhouse gases emissions. The ANTARES module produces 600 MWth which can be split into the required process heat, the remaining power drives an adapted prorated

  16. Nondestructive verification and assay systems for spent fuels. Technical appendixes

    International Nuclear Information System (INIS)

    Cobb, D.D.; Phillips, J.R.; Baker, M.P.

    1982-04-01

    Six technical appendixes are presented that provide important supporting technical information for the study of the application of nondestructive measurements to spent-fuel storage. Each appendix addresses a particular technical subject in a reasonably self-contained fashion. Appendix A is a comparison of spent-fuel data predicted by reactor operators with measured data from reprocessors. This comparison indicates a rather high level of uncertainty in previous burnup calculations. Appendix B describes a series of nondestructive measurements at the GE-Morris Operation Spent-Fuel Storage Facility. This series of experiments successfully demonstrated a technique for reproducible positioning of fuel assemblies for nondestructive measurement. The experimental results indicate the importance of measuring the axial and angular burnup profiles of irradiated fuel assemblies for quantitative determination of spent-fuel parameters. Appendix C is a reasonably comprehensive bibliography of reports and symposia papers on spent-fuel nondestructive measurements to April 1981. Appendix D is a compendium of spent-fuel calculations that includes isotope production and depletion calculations using the EPRI-CINDER code, calculations of neutron and gamma-ray source terms, and correlations of these sources with burnup and plutonium content. Appendix E describes the pulsed-neutron technique and its potential application to spent-fuel measurements. Although not yet developed, the technique holds the promise of providing separate measurements of the uranium and plutonium fissile isotopes. Appendix F describes the experimental program and facilities at Los Alamos for the development of spent-fuel nondestructive measurement systems. Measurements are reported showing that the active neutron method is sensitive to the replacement of a single fuel rod with a dummy rod in an unirradiated uranium fuel assembly

  17. PEMODELAN TERAS UNTUK ANALISIS PERHITUNGAN KONSTANTA MULTIPLIKASI REAKTOR HTR-PROTEUS

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-04-01

    Full Text Available PTRKN sebagai salah satu unit kerja di BATAN dengan tugas pokok dan fungsi yang berkaitan erat dengan teknologi reaktor dan keselamatan nuklir, menaruh perhatian khusus pada konsep reaktor pebble bed. Dalam makalah ini pemodelan reaktor pebble bed HTR-PROTEUS dilakukan dengan program transport Monte Carlo MCNP5. Partikel bahan bakar berlapis TRISO dimodelkan secara detail dan eksak dimana distribusi acak partikel ini dalam bola bahan bakar didekati menggunakan array teratur kisi SC dengan fraksi packing 5,76% tanpa zona eksklusif. Model teras pebble bed didekati dengan memanfaatkan kisi teratur dari bola yang disusun sebagai kisi BCC berdasarkan sel berulang yang digenerasi dari sejumlah sel satuan. Hasil perhitungan MCNP5 memperlihatkan kesesuaian yang sangat baik dengan eksperimen, walaupun teras HTR-PROTEUS diprediksi lebih reaktif daripada pengukuran, khususnya di teras 4.2 dan 4.3. Pustaka ENDF/B-VI menunjukkan konsistensi dengan estimasi keff paling akurat dibandingkan pustaka ENDF/B-V, terutama ENDF/B-VI (66c. Deviasi estimasi keff yang dihitung dengan eksperimen dikaitkan sebagai konsekuensi dari komposisi reflektor grafit yang dispesifikasikan. Komparasi yang dibuat memperlihatkan bahwa MCNP5 menghasilkan keff teras HTR-PROTEUS lebih presisi daripada hasil dari MCNP4B dan MCNPBALL. Hasil ini menyimpulkan bahwa, sukses metodologi pemodelan ini menjustifikasi aplikasi MCNP5 untuk analisis reaktor pebble bed lainnya. Kata kunci: pemodelan teras HTR-PROTEUS, konstanta multiplikasi, MCNP5   PTRKN as a working unit in BATAN whose main duties and functions are related to reactor technology and nuclear safety, consern attention to pebble bed reactor concept. In this paper modeling of HTR-PROTEUS pebble bed reactor was done using Monte Carlo transport code MCNP5. The TRISO coated fuel particle is modeled in detailed and exact manner where random distributions of these particles in fuel pebble is approximated by using regular array of SC lattice

  18. Spent fuel treatment in Japan

    International Nuclear Information System (INIS)

    Takahashi, K.

    1999-01-01

    In Japan, 52 nuclear power reactors are operating with a total power generation capacity of 45 GWe. The cumulative amount of spent fuel arising, as of March 1998, is about 14,700 W. Spent fuel is reprocessed and recovered nuclear materials are to be recycled in LWRs and FBRs. Pu utilization in LWRs will commence in 1999. In January 1997, short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of the reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, back-end measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away-from-reactor sites, considering the increasing amount of spent fuel arising. Valuable experience was been accumulated at the Tokai Reprocessing Plant (TRP), from the start of hot operation in 1977 up to now. The role of the TRP will be changed from an operation-oriented to a more R and D oriented facility, when PNC is reorganized into the new organization JNC. The Rokkasho reprocessing plant is under construction and is expected to commence operation in 2003. R and D of future recycling technologies is also continued for the establishment of a nuclear fuel cycle based on FBRs and LWRs. (author)

  19. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  20. Dry spent fuel storage licensing

    International Nuclear Information System (INIS)

    Sturz, F.C.

    1995-01-01

    In the US, at-reactor-site dry spent fuel storage in independent spent fuel storage installations (ISFSI) has become the principal option for utilities needing storage capacity outside of the reactor spent fuel pools. Delays in the geologic repository operational date at or beyond 2010, and the increasing uncertainty of the US Department of Energy's (DOE) being able to site and license a Monitored Retrievable Storage (MRS) facility by 1998 make at-reactor-site dry storage of spent nuclear fuel increasingly desirable to utilities and DOE to meet the need for additional spent fuel storage capacity until disposal, in a repository, is available. The past year has been another busy year for dry spent fuel storage licensing. The licensing staff has been reviewing 7 applications and 12 amendment requests, as well as participating in inspection-related activities. The authors have licensed, on a site-specific basis, a variety of dry technologies (cask, module, and vault). By using certified designs, site-specific licensing is no longer required. Another new cask has been certified. They have received one new application for cask certification and two amendments to a certified cask design. As they stand on the brink of receiving multiple applications from DOE for the MPC, they are preparing to meet the needs of this national program. With the range of technical and licensing options available to utilities, the authors believe that utilities can meet their need for additional spent fuel storage capacity for essentially all reactor sites through the next decade

  1. The expression and role of serotonin receptor 5HTR2A in canine osteoblasts and an osteosarcoma cell line.

    Science.gov (United States)

    Bracha, Shay; Viall, Austin; Goodall, Cheri; Stang, Bernadette; Ruaux, Craig; Seguin, Bernard; Chappell, Patrick E

    2013-12-12

    The significance of the serotonergic system in bone physiology and, more specifically, the importance of the five hydroxytryptamine receptor 2A (5HTR2A) in normal osteoblast proliferation have been previously described; however the role of serotonin in osteosarcoma remains unclear. Particularly, the expression and function of 5HTR2A in canine osteosarcoma has not yet been studied, thus we sought to determine if this indoleamine modulates cellular proliferation in vitro. Using real time quantitative reverse transcription PCR and immunoblot analyses, we explored receptor expression and signaling differences between non-neoplastic canine osteoblasts (CnOb) and an osteosarcoma cell line (COS). To elucidate specific serotonergic signaling pathways triggered by 5HTR2A, we performed immunoblots for ERK and CREB. Finally, we compared cell viability and the induction of apoptosis in the presence 5HTR2A agonists and antagonists. 5HTR2A was overexpressed in the malignant cell line in comparison to normal cells. In CnOb cells, ERK phosphorylation (ERK-P) decreased in response to both serotonin and a specific 5HTR2A antagonist, ritanserin. In contrast, ERK-P abundance increased in COS cells following either treatment. While endogenous CREB was undetectable in CnOb, CREB was observed constitutively in COS, with expression and exhibited increased CREB phosphorylation following escalating concentrations of ritanserin. To determine the influence of 5HTR2A signaling on cell viability we challenged cells with ritanserin and serotonin. Our findings confirmed that serotonin treatment promoted cell viability in malignant cells but not in normal osteoblasts. Conversely, ritanserin reduced cell viability in both the normal and osteosarcoma cells. Further, ritanserin induced apoptosis in COS at the same concentrations associated with decreased cell viability. These findings confirm the existence of a functional 5HTR2A in a canine osteosarcoma cell line. Results indicate that intracellular

  2. The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

    Energy Technology Data Exchange (ETDEWEB)

    Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

    2017-03-15

    Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

  3. Defining line replaceable units

    NARCIS (Netherlands)

    Parada Puig, J. E.; Basten, R. J I

    2015-01-01

    Defective capital assets may be quickly restored to their operational condition by replacing the item that has failed. The item that is replaced is called the Line Replaceable Unit (LRU), and the so-called LRU definition problem is the problem of deciding on which item to replace upon each type of

  4. Behavior of the P1.HTR mastocytoma cell line implanted in the chorioallantoic membrane of chick embryos

    Directory of Open Access Journals (Sweden)

    S.F. Avram

    2013-01-01

    Full Text Available The P1.HTR cell line includes highly transfectable cells derived from P815 mastocytoma cells originating from mouse breast tissue. Despite its widespread use in immunogenic studies, no data are available about the behavior of P1.HTR cells in the chick embryo chorioallantoic membrane model. The objective of the present investigation was to study the effects of P1.HTR cells implanted on the chorioallantoic membrane of chick embryos. We inoculated P1.HTR cells into the previously prepared chick embryo chorioallantoic membrane and observed the early and late effects of these cells by stereomicroscopy, histochemistry and immunohistochemistry. A highly angiotropic and angiogenic effect occurred early after inoculation and a tumorigenic potential with the development of mastocytoma keeping well mast cells immunophenotype was detected later during the development. The P1.HTR mastocytoma cell line is a good tool for the development of the chick embryo chorioallantoic membrane mastocytoma model and also for other studies concerning the involvement of blood vessels. The chick embryo chorioallantoic membrane model of mastocytoma retains the mast cell immunophenotype under experimental conditions and could be used as an experimental tool for in vivo preliminary testing of antitumor and antivascular drugs.

  5. Gene structure and expression of serotonin receptor HTR2C in hypothalamic samples from infanticidal and control sows

    Directory of Open Access Journals (Sweden)

    Quilter Claire R

    2012-04-01

    Full Text Available Abstract Background The serotonin pathways have been implicated in behavioural phenotypes in a number of species, including human, rat, mouse, dog and chicken. Components of the pathways, including the receptors, are major targets for drugs used to treat a variety of physiological and psychiatric conditions in humans. In our previous studies we have identified genetic loci potentially contributing to maternal infanticide in pigs, which includes a locus on the porcine X chromosome long arm. The serotonin receptor HTR2C maps to this region, and is therefore an attractive candidate for further study based on its function and its position in the genome. Results In this paper we describe the structure of the major transcripts produced from the porcine HTR2C locus using cDNA prepared from porcine hypothalamic and pooled total brain samples. We have confirmed conservation of sites altered by RNA editing in other mammalian species, and identified polymorphisms in the gene sequence. Finally, we have analysed expression and editing of HTR2C in hypothalamus samples from infanticidal and control animals. Conclusions The results confirm that although the expression of the long transcriptional variant of HTR2C is raised in infanticidal animals, the overall patterns of editing in the hypothalamus are similar between the two states. Sequences associated with the cDNA and genomic structures of HTR2C reported in this paper are deposited in GenBank under accession numbers FR720593, FR720594 and FR744452.

  6. Assessment of spent fuel cooling

    International Nuclear Information System (INIS)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

    1997-01-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD's work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools

  7. Analysis of the link between the redox state and enzymatic activity of the HtrA (DegP protein from Escherichia coli.

    Directory of Open Access Journals (Sweden)

    Tomasz Koper

    Full Text Available Bacterial HtrAs are proteases engaged in extracytoplasmic activities during stressful conditions and pathogenesis. A model prokaryotic HtrA (HtrA/DegP from Escherichia coli requires activation to cleave its substrates efficiently. In the inactive state of the enzyme, one of the regulatory loops, termed LA, forms inhibitory contacts in the area of the active center. Reduction of the disulfide bond located in the middle of LA stimulates HtrA activity in vivo suggesting that this S-S bond may play a regulatory role, although the mechanism of this stimulation is not known. Here, we show that HtrA lacking an S-S bridge cleaved a model peptide substrate more efficiently and exhibited a higher affinity for a protein substrate. An LA loop lacking the disulfide was more exposed to the solvent; hence, at least some of the interactions involving this loop must have been disturbed. The protein without S-S bonds demonstrated lower thermal stability and was more easily converted to a dodecameric active oligomeric form. Thus, the lack of the disulfide within LA affected the stability and the overall structure of the HtrA molecule. In this study, we have also demonstrated that in vitro human thioredoxin 1 is able to reduce HtrA; thus, reduction of HtrA can be performed enzymatically.

  8. Spent fuel element storage facility

    International Nuclear Information System (INIS)

    Ukaji, Hideo; Yamashita, Rikuo.

    1981-01-01

    Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)

  9. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  10. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Khan, A.; Pattantyus, P.

    1999-01-01

    The current status of the Canadian spent fuel storage is presented. This includes wet and dry interim storage. Extension of wet interim storage facilities is nor planned, as dry technologies have found wide acceptance. The Canadian nuclear program is sustained by commercial Ontario Hydro CANDU type reactors, since 1971, representing 13600 MW(e) of installed capacity, able to produce 9200 spent fuel bundles (1800 tU) every year, and Hydro Quebec and New Brunswick CANDU reactors each producing 685 MW(e) and about 100 tU of spent fuel annually. The implementation of various interim (wt and dry) storage technologies resulted in simple, dense and low cost systems. Economical factors determined that the open cycle option be adopted for the CANDU type reactors rather that recycling the spent fuel. Research and development activities for immobilization and final disposal of nuclear waste are being undertaken in the Canadian Nuclear Fuel Waste Management Program

  11. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Elder, H.K.

    1983-09-01

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate

  12. Transportation of spent MTR fuels

    International Nuclear Information System (INIS)

    Raisonnier, D.

    1997-01-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs

  13. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  14. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  15. HFIR spent fuel management alternatives

    International Nuclear Information System (INIS)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-01-01

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere

  16. Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Barnert-Wiemer, H.; Boehnert, R.

    1976-07-15

    The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

  17. Modulation of mitochondrial function and morphology by interaction of Omi/HtrA2 with the mitochondrial fusion factor OPA1

    Energy Technology Data Exchange (ETDEWEB)

    Kieper, Nicole; Holmstroem, Kira M.; Ciceri, Dalila; Fiesel, Fabienne C. [Center of Neurology and Hertie Institute for Clinical Brain Research, 72076 Tuebingen (Germany); Wolburg, Hartwig [Institute of Pathology, University of Tuebingen, 72076 Tuebingen (Germany); Ziviani, Elena; Whitworth, Alexander J. [Medical Research Council Centre for Developmental and Biomedical Genetics, University of Sheffield, Sheffield S10 2TN (United Kingdom); Martins, L. Miguel [Cell Death Regulation Laboratory, MRC Toxicology Unit, Leicester LE1 9HN (United Kingdom); Kahle, Philipp J., E-mail: philipp.kahle@uni-tuebingen.de [Center of Neurology and Hertie Institute for Clinical Brain Research, 72076 Tuebingen (Germany); Krueger, Rejko, E-mail: rejko.krueger@uni-tuebingen.de [Center of Neurology and Hertie Institute for Clinical Brain Research, 72076 Tuebingen (Germany)

    2010-04-15

    Loss of Omi/HtrA2 function leads to nerve cell loss in mouse models and has been linked to neurodegeneration in Parkinson's and Huntington's disease. Omi/HtrA2 is a serine protease released as a pro-apoptotic factor from the mitochondrial intermembrane space into the cytosol. Under physiological conditions, Omi/HtrA2 is thought to be involved in protection against cellular stress, but the cytological and molecular mechanisms are not clear. Omi/HtrA2 deficiency caused an accumulation of reactive oxygen species and reduced mitochondrial membrane potential. In Omi/HtrA2 knockout mouse embryonic fibroblasts, as well as in Omi/HtrA2 silenced human HeLa cells and Drosophila S2R+ cells, we found elongated mitochondria by live cell imaging. Electron microscopy confirmed the mitochondrial morphology alterations and showed abnormal cristae structure. Examining the levels of proteins involved in mitochondrial fusion, we found a selective up-regulation of more soluble OPA1 protein. Complementation of knockout cells with wild-type Omi/HtrA2 but not with the protease mutant [S306A]Omi/HtrA2 reversed the mitochondrial elongation phenotype and OPA1 alterations. Finally, co-immunoprecipitation showed direct interaction of Omi/HtrA2 with endogenous OPA1. Thus, we show for the first time a direct effect of loss of Omi/HtrA2 on mitochondrial morphology and demonstrate a novel role of this mitochondrial serine protease in the modulation of OPA1. Our results underscore a critical role of impaired mitochondrial dynamics in neurodegenerative disorders.

  18. Modulation of mitochondrial function and morphology by interaction of Omi/HtrA2 with the mitochondrial fusion factor OPA1

    International Nuclear Information System (INIS)

    Kieper, Nicole; Holmstroem, Kira M.; Ciceri, Dalila; Fiesel, Fabienne C.; Wolburg, Hartwig; Ziviani, Elena; Whitworth, Alexander J.; Martins, L. Miguel; Kahle, Philipp J.; Krueger, Rejko

    2010-01-01

    Loss of Omi/HtrA2 function leads to nerve cell loss in mouse models and has been linked to neurodegeneration in Parkinson's and Huntington's disease. Omi/HtrA2 is a serine protease released as a pro-apoptotic factor from the mitochondrial intermembrane space into the cytosol. Under physiological conditions, Omi/HtrA2 is thought to be involved in protection against cellular stress, but the cytological and molecular mechanisms are not clear. Omi/HtrA2 deficiency caused an accumulation of reactive oxygen species and reduced mitochondrial membrane potential. In Omi/HtrA2 knockout mouse embryonic fibroblasts, as well as in Omi/HtrA2 silenced human HeLa cells and Drosophila S2R+ cells, we found elongated mitochondria by live cell imaging. Electron microscopy confirmed the mitochondrial morphology alterations and showed abnormal cristae structure. Examining the levels of proteins involved in mitochondrial fusion, we found a selective up-regulation of more soluble OPA1 protein. Complementation of knockout cells with wild-type Omi/HtrA2 but not with the protease mutant [S306A]Omi/HtrA2 reversed the mitochondrial elongation phenotype and OPA1 alterations. Finally, co-immunoprecipitation showed direct interaction of Omi/HtrA2 with endogenous OPA1. Thus, we show for the first time a direct effect of loss of Omi/HtrA2 on mitochondrial morphology and demonstrate a novel role of this mitochondrial serine protease in the modulation of OPA1. Our results underscore a critical role of impaired mitochondrial dynamics in neurodegenerative disorders.

  19. Value-creating investment strategies to manage risk from structural market uncertainties: Switching and compound options in (V)HTR technologies - HTR2008-58157

    International Nuclear Information System (INIS)

    Lauferts, U.; Halbe, C.; Van Heek, A.

    2008-01-01

    To measure the value of a technology investment under uncertainty with standard techniques like net present value (NPV) or return on investment (ROI) will often uncover the difficulty to present convincing business case. Projected cash flows are inefficient or the discount rate chosen to compensate for the risk is so high, that it is disagreeable to the investor s requirements. Decision making and feasibility studies have to look beyond traditional analysis to reveal the strategic value of a technology investment. Here, a Real Option Analysis (ROA) offers a powerful alternative to standard discounted cash-flow (DCF) methodology by risk-adjusting the cash flow along the decision path rather than risk adjusting the discount rate. Within the GEN IV initiative attention is brought not only towards better sustainability, but also to broader industrial application and improved financing. Especially the HTR design is full of strategic optionalities: The high temperature output facilitates penetration into other non-electricity energy markets like industrial process heat applications and the hydrogen market. The flexibility to switch output in markets with multi-source uncertainties reduces downside risk and creates an additional value of over 50% with regard to the Net Present Value without flexibility. The supplement value of deploying a modular (V)HTR design adds over 100% to the project value using real option evaluation tools. Focus of this paper was to quantify the strategic value that comes along a) with the modular design; a design that offers managerial flexibility adapting a step-by-step investment strategy to the actual market demand and b) with the option to switch between two modes of operation, namely electricity and hydrogen production. We will demonstrate that the effect of uncertain electricity prices can be dampened down with a modular HTR design. By using a real option approach, we view the project as a series of compound options - each option depending

  20. Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

    2010-01-01

    Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

  1. Pilot study of dynamic Bayesian networks approach for fault diagnostics and accident progression prediction in HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunfei; Tong, Jiejuan; Zhang, Liguo, E-mail: lgzhang@tsinghua.edu.cn; Zhang, Qin

    2015-09-15

    Highlights: • Dynamic Bayesian network is used to diagnose and predict accident progress in HTR-PM. • Dynamic Bayesian network model of HTR-PM is built based on detailed system analysis. • LOCA Simulations validate the above model even if part monitors are lost or false. - Abstract: The first high-temperature-reactor pebble-bed demonstration module (HTR-PM) is under construction currently in China. At the same time, development of a system that is used to support nuclear emergency response is in progress. The supporting system is expected to complete two tasks. The first one is diagnostics of the fault in the reactor based on abnormal sensor measurements obtained. The second one is prognostic of the accident progression based on sensor measurements obtained and operator actions. Both tasks will provide valuable guidance for emergency staff to take appropriate protective actions. Traditional method for the two tasks relies heavily on expert judgment, and has been proven to be inappropriate in some cases, such as Three Mile Island accident. To better perform the two tasks, dynamic Bayesian networks (DBN) is introduced in this paper and a pilot study based on the approach is carried out. DBN is advantageous in representing complex dynamic systems and taking full consideration of evidences obtained to perform diagnostics and prognostics. Pearl's loopy belief propagation (LBP) algorithm is recommended for diagnostics and prognostics in DBN. The DBN model of HTR-PM is created based on detailed system analysis and accident progression analysis. A small break loss of coolant accident (SBLOCA) is selected to illustrate the application of the DBN model of HTR-PM in fault diagnostics (FD) and accident progression prognostics (APP). Several advantages of DBN approach compared with other techniques are discussed. The pilot study lays the foundation for developing the nuclear emergency response supporting system (NERSS) for HTR-PM.

  2. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1985-01-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

  3. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  4. Details of modelling HTR core physics: the use of pseudo nuclide traces

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Haas, J.B.M. de; Da Cruz, D.F.

    2003-01-01

    At present most combined neutronic and thermal hydraulic analyses of reactors, and the HTR is no exception, are being performed by codes employing few-group (typically 2-group) neutronics on the basis of parametrized few-group macroscopic (and microscopic) cross sections for homogenized areas, depending on quantities like irradiation (fuel only), 135 Xe concentration, temperature, etc. The irradiation parameter (time-integrated power per unit initial heavy metal mass) is sufficient for keeping track of the evolution of areas containing fuel. However, the use of the same parameter in areas without fuel, e.g. containing burnable poison, requires some special provisions. This can be met by the introduction of pseudo nuclides, with very specific cross sections and reaction chains, in the procedure to generate the parametrized few-group cross sections. It is shown that the time-evolution of a non-fuelled burnable poison area, as calculated by the 2-group (HTR) reactor code PANTHERMIX employing pseudo nuclides, compares well to the time-evolution obtained from an explicit burnup calculation by the WIMS8A/SNAP code. Examples are also shown using the pseudo nuclide method to keep track of the fast fluence (time-integrated flux above 0.1 MeV) in a continuous reload pebble-bed HTR reactor calculation by PANTHERMIX. Although the present implementation of the pseudo nuclide method exhibits some peculiarities connected to the specific codes in use (WIMS8A and PANTHERMIX) it is considered to be sufficiently general to be applicable in other code suites, requiring only limited modifications. (authors)

  5. Details of modelling HTR core physics: the use of pseudo nuclide traces

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Haas, J.B.M. de; Da Cruz, D.F. [Nuclear Research and consultancy Group (NRG), Petten (Netherlands)

    2003-07-01

    At present most combined neutronic and thermal hydraulic analyses of reactors, and the HTR is no exception, are being performed by codes employing few-group (typically 2-group) neutronics on the basis of parametrized few-group macroscopic (and microscopic) cross sections for homogenized areas, depending on quantities like irradiation (fuel only), {sup 135}Xe concentration, temperature, etc. The irradiation parameter (time-integrated power per unit initial heavy metal mass) is sufficient for keeping track of the evolution of areas containing fuel. However, the use of the same parameter in areas without fuel, e.g. containing burnable poison, requires some special provisions. This can be met by the introduction of pseudo nuclides, with very specific cross sections and reaction chains, in the procedure to generate the parametrized few-group cross sections. It is shown that the time-evolution of a non-fuelled burnable poison area, as calculated by the 2-group (HTR) reactor code PANTHERMIX employing pseudo nuclides, compares well to the time-evolution obtained from an explicit burnup calculation by the WIMS8A/SNAP code. Examples are also shown using the pseudo nuclide method to keep track of the fast fluence (time-integrated flux above 0.1 MeV) in a continuous reload pebble-bed HTR reactor calculation by PANTHERMIX. Although the present implementation of the pseudo nuclide method exhibits some peculiarities connected to the specific codes in use (WIMS8A and PANTHERMIX) it is considered to be sufficiently general to be applicable in other code suites, requiring only limited modifications. (authors)

  6. Software Unit Testing during the Development of Digital Reactor Protection System of HTR-PM

    International Nuclear Information System (INIS)

    Guo Chao; Xiong Huasheng; Li Duo; Zhou Shuqiao; Li Jianghai

    2014-01-01

    Reactor Protection System (RPS) of High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated in the Nuclear Power Plant (NPP) of China, and its development process has receives a lot of concerns around the world. As a 1E-level safety system, the RPS has to be designed and developed following a series of nuclear laws and technical disciplines including software verification and validation (software V&V). Software V&V process demonstrates whether all stages during the software development are performed correctly, completely, accurately, and consistently, and the results of each stage are testable. Software testing is one of the most significant and time-consuming effort during software V&V. In this paper, we give a comprehensive introduction to the software unit testing during the development of RPS in HTR-PM. We first introduce the objective of the testing for our project in the aspects of static testing, black-box testing, and white-box testing. Then the testing techniques, including static testing and dynamic testing, are explained, and the testing strategy we employed is also introduced. We then introduce the principles of three kinds of coverage criteria we used including statement coverage, branch coverage, and the modified condition/decision coverage. As a 1E-level safety software, testing coverage needs to be up to 100% mandatorily. Then we talk the details of safety software testing during software development in HTR-PM, including the organization, methods and tools, testing stages, and testing report. The test result and experiences are shared and finally we draw a conclusion for the unit testing process. The introduction of this paper can contribute to improve the process of unit testing and software development for other digital instrumentation and control systems in NPPs. (author)

  7. Utilization of heat from High Temperature Reactors (HTR) for dry reforming of methane

    Science.gov (United States)

    Jastrząb, Krzysztof

    2018-01-01

    One of the methods for utilization of waste carbon dioxide consists in reaction of methane with carbon dioxide, referred to as dry reforming of methane. It is an intensely endothermic catalytic process that takes place at the temperature above 700°C. Reaction of methane with carbon dioxide leads to formation of synthesis gas (syngas) that is a valuable chemical raw material. The energy that is necessary for the process to take place can be sourced from High Temperature Nuclear Reactors (HTR). The completed studies comprises a series of thermodynamic calculations and made it possible to establish optimum conditions for the process and demand for energy from HTR units. The dry reforming of methane needs also a catalytic agent with appropriate activity, therefore the hydrotalcite catalyser with admixture of cerium and nickel, developed at AGH University of Technology seems to be a promising solution. Thus, the researchers from the Institute for Chemical Processing of Coal (IChPW) in Zabrze have developed a methodology for production of the powdery hydrotalcite catalyser and investigated catalytic properties of the granulate obtained. The completed experiments confirmed that the new catalyser demonstrated high activity and is suitable for the process of methane dry reforming. In addition, optimum parameters of the were process (800°C, CO2:CH4 = 3:1) were established as well. Implementation of the technology in question into industrial practice, combined with utilization of HTR heat can be a promising method for management of waste carbon dioxide and may eventually lead to mitigation of the greenhouse effect.

  8. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    Energy Technology Data Exchange (ETDEWEB)

    Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

    1985-07-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

  9. Educating My Replacement

    Science.gov (United States)

    Tarter, Jill

    The search for extraterrestrial intelligence (SETI) could succeed tomorrow, decades from now, or never. The nature of this scientific exploration is such that we cannot predict success on any timescale; we only know that if we do not search, we cannot succeed. Having spent my scientific career in this field, I know perhaps better than anyone that the researchers of tomorrow may hold the key. Thus I have an enormous and vested interest in trying to educate the next generation of scientists. Because SETI excites such enthusiasm in young and old alike, I have an excellent opportunity to capture hearts and minds and leverage this interest into science education at many levels. Astrobiology is the new banner for inter- and cross-disciplinary investigations aimed at answering the big question "Are we alone?" The story of cosmic evolution is one that scientists at the SETI Institute have been telling for decades. We have used it as the framework for developing supplementary materials for elementary and middle schools called Life In The Universe. Currently we are tackling a year-long curriculum called Voyages Through Time for ninth grade students. This curriculum is delivered on CD-ROM and supported by the web. It focuses on evolution as a theme and stresses the contributions made from all the traditionally isolated branches of science --- and by the way, it's fun! I am a product of the post-Sputnik era and the American emphasis on science and engineering education. In the New York City bedroom community where I grew up, every school bond issue passed at every election. So I am appalled at the difficulties, the impecuniousness, and bureaucratic nonsense our pilot and field test teachers encounter on a daily basis. I am also overjoyed that even under such unreasonable conditions, I meet enthusiastic teachers who care about their students and are dedicated to helping them achieve the best possible education. Not all students will become scientists, nor should they. However

  10. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  11. HTR combustion head end comparison of the shaft furnace and fluidized bed processes

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Kaiser, G.; Pirk, H.; Tillessen, U.

    1975-01-15

    Two methods are described for the combustion of the graphite of HTR fuel elements, a sufficient description of the principles being given to permit an understanding of the processes. The present state of the technology of the two processes is then compared on the basis of the results obtained at Gulf General Atomic. Finally, the possibilities of further development are examined using a pilot plant designed to deliver a reactor power of 7000 MWe as the basis. The present report is a collection of facts. It contains neither an evaluation nor a recommendation. A summarized comparison of the state of the technology and the possibilities of development is given in tabular form.

  12. Distribution of tritium in a nuclear process heat plant with HTR

    International Nuclear Information System (INIS)

    Steinwarz, W.; Stoever, D.; Hecker, R.; Thiele, W.

    1984-01-01

    The application of HTR-process heat in chemical processes involves low contamination of the product by tritium permeation through the heat exchanger walls. According to conservative assumptions for the tritium release rate and based on experimental permeation data of the German R und D-program a tritium concentration in the PNP-product gas of about 10 pCi/g was calculated. The domestic use of the product gas in unvented kitchen ranges as the most important direct radiation exposure pathway then leads to an effective equivalent radiation dose of only 20 μrem/a. (orig.)

  13. Fission product behaviour - in particular Cs-137 - in HTR-TRISO-coated particle fuel

    International Nuclear Information System (INIS)

    Allelein, H.J.

    1980-12-01

    This work is performed between 1977 and 1979. The main task is to determine a temperature dependent diffusion coefficient of the fission product Cs-137 in the silicon carbide interlayer of HTR particles. The raw material is laso presented as the used measuring techniques and computer codes. The results are discussed in detail and some critical remarks are made about the efficiency of the silicon carbide interlayer to retent fission products including Ag-110m, Sr-90, and Ru-106, which temperature dependent diffusion coefficient is also been determined. (orig.) [de

  14. The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

    2006-07-15

    This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

  15. Key technology for (V)HTR: laser beam joining of SiC

    International Nuclear Information System (INIS)

    Knorr, J.; Lippmann, W.; Reinecke, A.M.; Wolf, R.; Rasper, R.; Kerber, A.; Wolter, A.

    2005-01-01

    Laser beam joining has numerous advantages over other methods presently known. After having been developed successful for brazing silicon carbide for high temperature applications, this technology is now also available for silicon nitride. Thus the field of application of SiC and Si 3 N 4 which are very interesting materials for the nuclear sector is considerably extended thanks to this new technology. Ceramic encapsulation of fuel and absorber increases the margins for operation at very high temperatures. Additionally, without ceramic encapsulation of the main core components, it will be difficult to continue claiming non-catastrophic behaviour for the (V)HTR. (orig.)

  16. Results of a German probabilistic risk assessment study for the HTR-1160 concept

    International Nuclear Information System (INIS)

    Fassbender, J.; Kroeger, W.

    1981-01-01

    The paper reviews ''Accident Initiations and Progression Analysis'' methodology and results which applied to the German equivalent of the HTGR-1160 and German site conditions. The investigation of accidents contributing to risk was concentrated on those event sequences which lead to major release of core inventory or - with less importance - to release of plate-out activity together with coolant gas activity. With regard to release mechanisms severe HTR-accidents were grouped into: a) water ingress events with fission product release due to hydrolysis of defective coated particles and desorption of plate out activity and b) core heating events with fission product release after coated particle failure due to excessive temperatures

  17. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  18. HTR 500: Final report of the project '' uniaxial creep tests at controlled temperature''

    International Nuclear Information System (INIS)

    1992-03-01

    The report presents the results of creep trials with HTR-concrete, which were carried out in the scope of development of prestressed concrete - reactor pressure vessels at the ETH Lausanne Institute for Steel and Prestressed Concrete. With temperature, an increase of creep and shrinkage was observed, a lesser dependence on exhaustion and type of concrete. The point in time of reaching the final value is not dependent on temperature for creep, but is for shrinkage. The modulus of elasticity depends on the temperature pre-treatment, but only insignificantly on the test temperature. figs., tabs

  19. Research on vibration properties of auxiliary bearing cage used in HTR-10 GT project

    International Nuclear Information System (INIS)

    Qin Qingquan; Yang Guojun; Shi Zhengang; Yu Suyuan

    2009-01-01

    Auxiliary Bearings (ABs) is one of the most important parts in Active Magnetic Bearing (AMB) system, which was used in HTR-10 GT project. This paper uses finite element method to analyze the centrifugal stress and free vibration properties of the cage according to its work condition. And different geometric parameters of the cage that has effects on its vibration performance are discussed. The results show that the highest centrifugal stress is in the middle of the cage side sill. The low odder vibration modes of the cage can be induced when the auxiliary bearings are working. Proper geometric parameters and ball pocket number can enhance the performance of the cage. (authors)

  20. Results of a German probabilistic risk assessment study for the HTR-1160 concept

    Energy Technology Data Exchange (ETDEWEB)

    Fassbender, J.; Kroeger, W. [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Nukleare Sicherheitsforschung

    1981-01-15

    The paper reviews ''Accident Initiations and Progression Analysis'' methodology and results which applied to the German equivalent of the HTGR-1160 and German site conditions. The investigation of accidents contributing to risk was concentrated on those event sequences which lead to major release of core inventory or - with less importance - to release of plate-out activity together with coolant gas activity. With regard to release mechanisms severe HTR-accidents were grouped into: a) water ingress events with fission product release due to hydrolysis of defective coated particles and desorption of plate out activity and b) core heating events with fission product release after coated particle failure due to excessive temperatures.

  1. Technology Assessment HTR. Part 7. Social support for the introduction of the High Temperature Reactor

    International Nuclear Information System (INIS)

    De Ruiter, W.

    1996-06-01

    The safety of nuclear power plants is the main subject in risk analysis and risk perception of nuclear energy. The question is if a substantial increase in safety according to the classic risk analysis will lead to a decrease in the percepted risks of nuclear energy. In this report the uncertainties in existing risk analysis are dealt with. The results of public risk perception studies of nuclear power are then analysed and possible changes in the public risk perception in the case of the introduction of the HTR is dealt with. Results and conclusions are presented. 4 tabs., 71 refs

  2. Pu and MA Management in Thermal HTR, QUO VADIS? Insights from the Euratom PUMA project

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    2013-01-01

    The results of this study demonstrate the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burn-up of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burn-up and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the “wallpaper” fuel does not have advantage over the standard fuel design in this respect

  3. Ankle replacement - discharge

    Science.gov (United States)

    ... total - discharge; Total ankle arthroplasty - discharge; Endoprosthetic ankle replacement - discharge; Osteoarthritis - ankle ... You had an ankle replacement. Your surgeon removed and reshaped ... an artificial ankle joint. You received pain medicine and were ...

  4. Artificial Disc Replacement

    Science.gov (United States)

    ... Spondylolisthesis BLOG FIND A SPECIALIST Treatments Artificial Disc Replacement (ADR) Patient Education Committee Jamie Baisden The disc ... Disc An artificial disc (also called a disc replacement, disc prosthesis or spine arthroplasty device) is a ...

  5. Partial knee replacement - slideshow

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/presentations/100225.htm Partial knee replacement - series—Normal anatomy To use the sharing ... A.M. Editorial team. Related MedlinePlus Health Topics Knee Replacement A.D.A.M., Inc. is accredited ...

  6. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  7. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

    The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

  8. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Weiner, R.F.

    1991-01-01

    As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel

  9. Flued head replacement alternatives

    International Nuclear Information System (INIS)

    Smetters, J.L.

    1987-01-01

    This paper discusses flued head replacement options. Section 2 discusses complete flued head replacement with a design that eliminates the inaccessible welds. Section 3 discusses alternate flued head support designs that can drastically reduce flued head installation costs. Section 4 describes partial flued head replacement designs. Finally, Section 5 discusses flued head analysis methods. (orig./GL)

  10. Capital Equipment Replacement Decisions

    OpenAIRE

    Batterham, Robert L.; Fraser, K.I.

    1995-01-01

    This paper reviews the literature on the optimal replacement of capital equipment, especially farm machinery. It also considers the influence of taxation and capital rationing on replacement decisions. It concludes that special taxation provisions such as accelerated depreciation and investment allowances are unlikely to greatly influence farmers' capital equipment replacement decisions in Australia.

  11. Implementing Replacement Cost Accounting

    Science.gov (United States)

    1976-12-01

    cost accounting Clickener, John Ross Monterey, California. Naval Postgraduate School http://hdl.handle.net/10945/17810 Downloaded from NPS Archive...Calhoun IMPLEMENTING REPLACEMENT COST ACCOUNTING John Ross CHckener NAVAL POSTGRADUATE SCHOOL Monterey, California THESIS IMPLEMENTING REPLACEMENT COST ...Implementing Replacement Cost Accounting 7. AUTHORS John Ross Clickener READ INSTRUCTIONS BEFORE COMPLETING FORM 3. RECIPIENT’S CATALOG NUMBER 9. TYRE OF

  12. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  13. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

    1997-12-01

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  14. Needs in Research and Development on materials for the gas coolant nuclear system: HTR/VHTR and GFR; Besoins en R et D sur les materiaux pour les systemes nucleaires a caloporteur gaz: HTR/VHTR et GFR

    Energy Technology Data Exchange (ETDEWEB)

    Billot, Ph. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France)

    2003-07-01

    This presentation takes stock on the materials for high temperature reactors HTR (850 C), very high temperature VHTR(>1000 C) and fast neutrons high temperature GGF(850 C). It concerns the welding materials for the vessel, Ni-based superalloys for gas turbines, coatings, graphite, ceramics and corrosion studies. (A.L.B.)

  15. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Shirahashi, K.; Maeda, M.; Nakai, T.

    1996-01-01

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  16. Spent Pot Lining Characterization Framework

    Science.gov (United States)

    Ospina, Gustavo; Hassan, Mohamed I.

    2017-09-01

    Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.

  17. Spent fuel storage requirements, 1988

    International Nuclear Information System (INIS)

    1988-10-01

    Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs

  18. Recycling of waste spent catalyst in road construction and masonry blocks.

    Science.gov (United States)

    Taha, Ramzi; Al-Kamyani, Zahran; Al-Jabri, Khalifa; Baawain, Mahad; Al-Shamsi, Khalid

    2012-08-30

    Waste spent catalyst is generated in Oman as a result of the cracking process of petroleum oil in the Mina Al-Fahl and Sohar Refineries. The disposal of spent catalyst is of a major concern to oil refineries. Stabilized spent catalyst was evaluated for use in road construction as a whole replacement for crushed aggregates in the sub-base and base layers and as a partial replacement for Portland cement in masonry blocks manufacturing. Stabilization is necessary as the waste spent catalyst exists in a powder form and binders are needed to attain the necessary strength required to qualify its use in road construction. Raw spent catalyst was also blended with other virgin aggregates, as a sand or filler replacement, for use in road construction. Compaction, unconfined compressive strength and leaching tests were performed on the stabilized mixtures. For its use in masonry construction, blocks were tested for unconfined compressive strength at various curing periods. Results indicate that the spent catalyst has a promising potential for use in road construction and masonry blocks without causing any negative environmental impacts. Copyright © 2012 Elsevier B.V. All rights reserved.

  19. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gago, J.A.; Gravalos, J.M.

    1996-01-01

    There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

  20. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  1. Transport device of spent fuel

    International Nuclear Information System (INIS)

    Watanabe, Takashi.

    1976-01-01

    Object: To provide a transport device of spent fuel particularly used in a fast breeder, which can enhance accessibility to travelling mechanism portions and exchangeability thereof to facilitate maintenance in the event of failure. Structure: On a travelling floor, which has a function to shield radioactive rays, extending in a direction of transporting spent fuel and being formed with a break passing through in a direction wall thickness, a travelling body is moved along the break. The travelling body has a support rod member mounted thereon, and the support rod member is moved within the break, the support rod member having a fuel support pocket suspended therefrom. (Furukawa, Y.)

  2. DEM simulation of particle mixing for optimizing the overcoating drum in HTR fuel fabrication

    Science.gov (United States)

    Liu, Malin; Lu, Zhengming; Liu, Bing; Shao, Youlin

    2013-06-01

    The rotating drum was used for overcoating coated fuel particles in HTR fuel fabrication process. All the coated particles should be adhered to equal amount of graphite powder, which means that the particle should be mixed quickly in both radial and axial directions. This paper investigated the particle flow dynamics and mixing behavior in different regimes using the discrete element method (DEM). By varying the rotation speed, different flow regimes such as slumping, rolling, cascading, cataracting, centrifuging were produced. The mixing entropy based on radial and axial grid was introduced to describe the radial and axial mixing behaviors. From simulation results, it was found that the radial mixing can be achieved in the cascading regime more quickly than the slumping, rolling and centrifuging regimes, but the traditional rotating drum without internal components can not achieve the requirements of axial mixing and should be improved. Three different structures of internal components are proposed and simulated. The new V-shaped deflectors were found to achieve a quick axial mixing behavior and uniform axial distribution in the rotating drum based on simulation results. At last, the superiority was validated by experimental results, and the new V-shaped deflectors were used in the industrial production of the overcoating coated fuel particles in HTR fuel fabrication process.

  3. The improvement of the method of equivalent cross section in HTR

    International Nuclear Information System (INIS)

    Guo, J.; Li, F.

    2012-01-01

    The Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusion method. By appropriately adjusting the diffusion coefficient of homogenized absorber region, the diffusion theory could yield satisfactory results for the full core model with strong neutron absorber material, for example the control rod in High temperature gas cooled reactor (HTR). Original implementation of MECS based on 1-D cell transport model has some limitation on accuracy and applicability, a new implementation of MECS based on 2-D transport model are proposed and tested in this paper. This improvement can extend the MECS to the calculation of twin small absorber ball system which have a non-circular boring in graphite reflector and different radial position. A least-square algorithm for the calculation of equivalent diffusion coefficient is adopted, and special treatment for diffusion coefficient for higher energy group is proposed in the case that absorber is absent. Numerical results to adopt MECS into control rod calculation in HTR are encouraging. However, there are some problems left. (authors)

  4. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    International Nuclear Information System (INIS)

    Yi, DU; Xiangang, WANG; Xincheng, XIANG; Bing, LIU

    2014-01-01

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production

  5. Safety related studies on the accident behaviour of the HTR-100

    International Nuclear Information System (INIS)

    Wolters, J.; Mertens, J.; Altes, J.; Bongartz, R.; Breitbach, G.; David, P.H.; Degen, G.; Ehrlich, H.G.; Escherich, K.H.; Frank, E.; Hennings, W.; Jahn, W.; Koschmieder, R.; Marx, J.; Meister, G.; Moormann, R.; Rehm, W.; Verfondern, K.

    1991-10-01

    The aim of investigations was to verify the safety concept of the plant for balance and to quantify the radiological risk to be expected in operating an HTR-100 double unit system. Moreover, aspects of the investment risk were considered. The spectrum of initiating events ranged from so-called transients to leaks in the primary circuit and steam generator and even included earthquakes. Some of the event trees derived were highly complex and extensive due to the situation of the steam generator above the core and with regard to the double unit plant concept with increased possibilities of accident control, but also with respect to potential accident propagation. Correspondingly sophisticated analyses were required to identify risk-relevant event sequences. Environmental exposure for all risk-relevant accidents is so low that accident consequence calculations do not reveal any lethal radiation doses and practically no stochastic fatal injuries. These calculations neither assumed acute protective measures nor long-term resettlement or decontamination. The radiological risk caused by an HTR-100 plant is therefore to be classified as very low. The initiating events selected as representative and the event sequences studied in detail cover the risk-relevant event spectrum well into the hypothetical range. (orig./HP) [de

  6. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  7. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Yi, DU, E-mail: duyi11@mails.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building A309, Haidian District, Beijing 100084 (China); Xiangang, WANG, E-mail: wangxiangang@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building A309, Haidian District, Beijing 100084 (China); Xincheng, XIANG, E-mail: inetxxc@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building, Haidian District, Beijing 100084 (China); Bing, LIU, E-mail: bingliu@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building, Haidian District, Beijing 100084 (China)

    2014-12-15

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production.

  8. Analysis of Seismic Soil-Structure Interaction for a Nuclear Power Plant (HTR-10

    Directory of Open Access Journals (Sweden)

    Xiaoxin Wang

    2017-01-01

    Full Text Available The response of nuclear power plants (NPPs to seismic events is affected by soil-structure interactions (SSI. In the present paper, a finite element (FE model with transmitting boundaries is used to analyse the SSI effect on the response of NPP buildings subjected to vertically incident seismic excitation. Analysis parameters that affect the accuracy of the calculations, including the dimension of the domain and artificial boundary types, are investigated through a set of models. A numerical SSI analysis for the 10 MW High Temperature Gas Cooled Test Reactor (HTR-10 under seismic excitation was carried out using the developed model. The floor response spectra (FRS produced by the SSI analysis are compared with a fixed-base model to investigate the SSI effect on the dynamic response of the reactor building. The results show that the FRS at foundation level are reduced and those at higher floor levels are altered significantly when taking SSI into account. The peak frequencies of the FRS are reduced due to the SSI, whereas the acceleration at high floor levels is increased at a certain frequency range. The seismic response of the primary system components, however, is reduced by the analysed SSI for the HTR-10 on the current soil site.

  9. A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

    Energy Technology Data Exchange (ETDEWEB)

    Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2013-06-15

    HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

  10. Prediction calculation of HTR-10 fuel loading for the first criticality

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

    2001-01-01

    The 10 MW high temperature gas cooled reactor (HTR-10) was built at Institute of Nuclear Energy Technology, Tsinghua University, and the first criticality was attained in Dec. 2000. The high temperature gas cooled reactor physics simulation code VSOP was used for the prediction of the fuel loading for HTR-10 first criticality. The number of fuel element and graphite element was predicted to provide reference for the first criticality experiment. The prediction calculations toke into account the factors including the double heterogeneity of the fuel element, buckling feedback for the spectrum calculation, the effect of the mixture of the graphite and the fuel element, and the correction of the diffusion coefficients near the upper cavity based on the transport theory. The effects of impurities in the fuel and the graphite element in the core and those in the reflector graphite on the reactivity of the reactor were considered in detail. The first criticality experiment showed that the predicted values and the experiment results were in good agreement with little relative error less than 1%, which means the prediction was successful

  11. Study of the Effect of Burnable Poison Particles Applying in a Pebble Bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing

    2014-01-01

    In pebble bed high temperature gas cooled reactors (HTR), spherical fuel elements pass through the core several times to balance the burnup process in the fuel region, resulting in an acceptable shape and peak factor of power density in the simulation analysis. In contrast, when fuel elements pass through the core only once, the peak of power density occurs at the top of the core and its value is too high to be safe. These indicators/parameters can be improved by incorporating burnable poison in the fuel elements under certain conditions. In the current study, burnable poison particles (BPPs) in fuel elements are evaluated. In spite of the strong absorption capability of "1"0B, BPPs can decrease the depletion speed and increase the duration of "1"0B because of the self-shielding effect, resulting in improved shape and peak factor of power distribution. Several BPPs with different radius are discussed in power distribution, following the calculation for a full-scale reactor core with modified VSOP code. According the result, applying BPPs on fuel pebbles is an effective means to improve the distribution of the power density under one-through fuel load in HTR. (author)

  12. Auxiliary bearing design and rotor dynamics analysis of blower fan for HTR-10

    International Nuclear Information System (INIS)

    Gao Mingshan; Yang Guojun; Xu Yang; Zhao Lei; Yu Suyuan

    2005-01-01

    The electromagnetic bearing instead of ordinary mechanical bearing was chosen to support the rotor in the blower fan system with helium of 10 MW high temperature gas-cooled test reactor (HTR-10), and the auxiliary bearing was applied in the HTR-10 as the backup protector. When the electromagnetic bearing doesn't work suddenly for the power broken, the auxiliary bearing is used to support the falling rotor with high rotating speed. The rotor system will be protected by the auxiliary bearing. The design of auxiliary bearing is the ultimate safeguard for the system. This rotor is vertically mounted to hold the blower fan. The rotor's length is about 1.5 m, its weight is about 240 kg and the rotating speed is about 5400 r/min. Auxiliary bearing design and rotor dynamics analysis are very important for the design of blower fan to make success. The research status of the auxiliary bearing was summarized in the paper. A sort of auxiliary bearing scheme was proposed. MSC.Marc was selected to analyze the vibration mode and the natural frequency of the rotor. The scheme design of auxiliary bearing and analysis result of rotor dynamics offer the important theoretical base for the protector design and control system of electromagnetic bearing of the blower fan. (authors)

  13. Coordinated Control Design for the HTR-PM Plant: From Theoretic Analysis to Simulation Verification

    International Nuclear Information System (INIS)

    Dong Zhe; Huang Xiaojin

    2014-01-01

    HTR-PM plant is a two-modular nuclear power plant based on pebble bed modular high temperature gas-cooled reactor (MHTGR), and adopts operation scheme of two nuclear steam supplying systems (NSSSs) driving one turbine. Here, an NSSS is composed of an MHTGR, a once-through steam generator (OTSG) and some connecting pipes. Due to the coupling effect induced by two NSSSs driving one common turbine and that between the MHTGR and OTSG given by common helium flow, it is necessary to design a coordinated control for the safe, stable and efficient operation of the HTR-PM plant. In this paper, the design of the feedback loops and control algorithms of the coordinated plant control law is firstly given. Then, the hardware-in-loop (HIL) system for verifying the feasibility and performance of this control strategy is introduced. Finally, some HIL simulation results are given, which preliminarily show that this coordinated control law can be implemented practically. (author)

  14. A network-based system of simulation, control and online assistance for HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Zhu Shutang [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)], E-mail: zhust@tsinghua.edu.cn; Luo Shaojie; Shi Lei [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2008-07-15

    A network-based computer system has been developed for HTR-10. This system integrates three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four functional elements: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate with each other via network. It is intended to analyze and calculate physical processes of the reactor core, the main loop system and the steam generator, etc., as well as to simulate the normal operational and transient accidents. The result data can be dynamically displayed through the RGDC. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameters, which are difficult to measure. This integrated system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online support for operators in the main control room, or as a convenient powerful tool for the control system design.

  15. A network-based system of simulation, control and online assistance for HTR-10

    International Nuclear Information System (INIS)

    Zhu Shutang; Luo Shaojie; Shi Lei

    2008-01-01

    A network-based computer system has been developed for HTR-10. This system integrates three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four functional elements: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate with each other via network. It is intended to analyze and calculate physical processes of the reactor core, the main loop system and the steam generator, etc., as well as to simulate the normal operational and transient accidents. The result data can be dynamically displayed through the RGDC. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameters, which are difficult to measure. This integrated system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online support for operators in the main control room, or as a convenient powerful tool for the control system design

  16. Spent fuel storage rack for atomic power plant

    International Nuclear Information System (INIS)

    Kodama, Tatemitsu.

    1981-01-01

    Purpose: To flexibly cope with the changes in the size and shape of spent fuel storage containers by placing a number of independently-constructed rack cells in a rack frame in such a manner that the guide support members of the storage rack, mounted on each rack cell may be replaced. Constitution: Independently-constructed rack cells are inserted from above into a rack frame rigidly installed on the bottom of a water pool. Each cell is produced by welding, has a handling head mounted at the top, and guide support members made of three replaceable guide tubes are mounted with bolts. If the size and the shape of the containers are altered, this configuration can easily cope with the new container shape by merely having the guide tubes replaced, without adversely affecting other cells and without necessitating draining of the water in the pool. (Yoshino, Y.)

  17. The development on-line monitoring system of active magnetic bearings for HTR-10GT

    International Nuclear Information System (INIS)

    Shi Zhengang; Shi Lei; Zha Meisheng; Yu Suyuan

    2005-01-01

    High Temperature Gas-cooled Reactor (HTR) is recognized as an advanced type of reactor incorporating many design enhancements such as inherent safety features, fuel cycle flexibility, highly fuel utilization, highly efficient electricity generation and process heat application. The research and development of HTR started at the middle of the 1970's, and came to be a part of the Chinese High Technology Program in 1986. A plan to build a 10 MW High Temperature Gas-cooled Reactor (HTR-10) was approved by the State Science and Technology Commission in 1990, and in 1995 the construction was initiated at the Institute of Nuclear Energy Technology (INET), Tsinghua University. The full power 10 MW operation for 72 hours have reached in 2003, and have been checked and accepted by the State Science and Technology Commission. In order to advance the HTR-10 performance, the project of the Helium Gas Turbine Generator for the HTR-10 was authorized by the State Science and Technology Commission, and stared in 2003. In this project, active magnetic bearings (AMBs) are chosen to support the generator rotor and the turbocompressor rotor in the power conversion unit because of their numerous advantages over the conventional bearings. In order to detect how the AMB system works in operation and make diagnosis whether the system behaves normally or not, the monitoring system based on the virtual instruments is designed to monitor the working conditions of the PCU, and to ensure its normal operation. This monitoring system consists of the industry personal computer (PC), the data acquisition system, the measurement transmitters and the LabVIEW system platform. It is located at the PCU control room, and communicates with the master control room by Controller Area Net (CAN). The development is divided into the following three steps: First, a data acquisition platform to collect and acquire all the necessary and useful data from the operation of the AMB system is developed. Second, the

  18. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K.; Cho, M. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW{sub th} block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW{sub th} block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared.

  19. Gadd45 α expression in preeclampsia placenta and the effect of Gadd45 α on trophoblast HTR8/Svneo

    Directory of Open Access Journals (Sweden)

    Li Wang

    2016-01-01

    Full Text Available Objective: To study the expression of Gadd45 α in preeclampsia placenta and the regulating effect of Gadd45 α knockdown on trophoblast HTR8/Svneo. Methods: Preeclampsia placenta tissue and normal placenta tissue were collected, and mRNA contents and protein contents of Gadd45 α were detected by fluorescent quantitative PCR and Western blotting respectively; trophoblast cells HTR8/Svneo were cultured and after transfection of Gadd45 α siRNA, cell invasion ability and expression of invasion-assiotiated molecules were detected. Results: mRNA content and protein content of Gadd45 α in preeclampsia placenta tissue were higher than those in normal placenta tissue; after transfection of Gadd45 α siRNA, mRNA content and protein content of Gadd45 α in HTR8/Svneo cells significantly decreased, and the number of invasive cells as well as expression of MMP1, MMP2, MMP3 and MMP9 significantly increased. Conclusion: The expression of Gadd45 α in preeclampsia placenta abnormally increases; inhibting the expression of Gadd45 α in trophoblasts HTR8/Svneo can promote invasion and increase the expression of MMPs molecules.

  20. Pre-economic analysis of HTR in preparation for a comprehensive economic assessment of HTRs in the world

    Energy Technology Data Exchange (ETDEWEB)

    Bredimas, Alexandre, E-mail: alexandre.bredimas@strane-innovation.com

    2014-05-01

    High temperature nuclear reactors will address mainly the industrial cogeneration market and compete with gas cogeneration, the current reference technology. The key question for HTR is therefore: how far are HTRs competitive against gas technologies? This simple question demands a complex response. First, the cogeneration scheme has to be discussed according the specificities in heat usage of every industry as they will impact the design. Second, the costs, revenues and risks of the different lifecycle phases for both a HTR and gas cogeneration plant have to be assessed and compared. These parameters will greatly depend on each location (personnel costs, gas local prices, CO{sub 2} pricing, etc.). A particular attention has to be given to the risk interactions between the cogeneration plant and the industrial facility it is supplying with heat and electricity (e.g. tritium contamination in industrial processes, explosion of flammable products in industrial site). This paper aims mainly at starting exchanges at international level with other equivalent initiatives in order to assess in general terms the economic viability of HTR worldwide, in relation to the evaluation of the HTR global market.

  1. Needs in Research and Development on materials for the gas coolant nuclear system: HTR/VHTR and GFR

    International Nuclear Information System (INIS)

    Billot, Ph.

    2003-01-01

    This presentation takes stock on the materials for high temperature reactors HTR (850 C), very high temperature VHTR(>1000 C) and fast neutrons high temperature GGF(850 C). It concerns the welding materials for the vessel, Ni-based superalloys for gas turbines, coatings, graphite, ceramics and corrosion studies. (A.L.B.)

  2. Pre-economic analysis of HTR in preparation for a comprehensive economic assessment of HTRs in the world

    International Nuclear Information System (INIS)

    Bredimas, Alexandre

    2014-01-01

    High temperature nuclear reactors will address mainly the industrial cogeneration market and compete with gas cogeneration, the current reference technology. The key question for HTR is therefore: how far are HTRs competitive against gas technologies? This simple question demands a complex response. First, the cogeneration scheme has to be discussed according the specificities in heat usage of every industry as they will impact the design. Second, the costs, revenues and risks of the different lifecycle phases for both a HTR and gas cogeneration plant have to be assessed and compared. These parameters will greatly depend on each location (personnel costs, gas local prices, CO 2 pricing, etc.). A particular attention has to be given to the risk interactions between the cogeneration plant and the industrial facility it is supplying with heat and electricity (e.g. tritium contamination in industrial processes, explosion of flammable products in industrial site). This paper aims mainly at starting exchanges at international level with other equivalent initiatives in order to assess in general terms the economic viability of HTR worldwide, in relation to the evaluation of the HTR global market

  3. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    International Nuclear Information System (INIS)

    Kim, Young Min; Jo, C. K.; Cho, M. S.

    2016-01-01

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW_t_h block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW_t_h block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared

  4. Process heat applications of HTR-PM600 in Chinese petrochemical industry: Preliminary study of adaptability and economy

    International Nuclear Information System (INIS)

    Fang, Chao; Min, Qi; Yang, Yanran; Sun, Yuliang

    2017-01-01

    Highlights: •High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. •The joint of a 600 MW modular HTGR (HTR-PM600) and petrochemical industry is achievable. •The mature technology of turbine in thermal power station could be readily adopted. •The economy of this scheme is also acceptable. -- Abstract: High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. In this article, the preliminary feasibility of a 600 MW modular HTGR (HTR-PM600) working as heat source for a typical hypothetical Chinese petrochemical factory is discussed and it is found that the joint of HTR-PM600 and petrochemical industry is achievable. In detail, the heat and water balance analysis of the petrochemical factory is given. Furthermore, the direct cost of heat supplied by HTR-PM600 is calculated and corresponding economy is estimated. The results show that though there are several challenges, the application of process heat of HTGR to petrochemical industry is practical in sense of both technology and economy.

  5. Expanded spent fuel storage project at Yankee Atomic Electric Plant

    International Nuclear Information System (INIS)

    Chin, S.L.

    1980-01-01

    A detailed discussion on the project at the Yankee Rowe power reactor for expanding the capacity of the at-reactor storage pool by building double-tier storage racks. Various alternatives for providing additional capacity were examined by the operators. Away-from-reactor alternatives included shipment to existing privately owned facilities, a regional independent storage facility, and transshipments to other New England nuclear power plant pools. At-reactor alternatives evaluated included a new pool modification of the existing structure and finally, modification of the spent fuel pit. The establishment of a federal policy precluding transshipment of spent fuel prohibited the use of off-site alternatives. The addition of another pool was too expensive. The possibility of modifying an existing on-site structure required a new safety evaluation by the regulatory group with significant cost and time delays. Therefore, the final alternative - utilizing the existing spent fuel pool with some modification - was chosen due to cost, licensing possibility, no transport requirements, and the fact that the factors involved were mainly under the control of the operator. Modification of the pool was accomplished in phases. In the first phase, a dam was installed in the center of the pool (after the spent fuel was moved to one end). In the second phase, the empty end of the pool was drained and lined with stainless steel and the double-tier rack supports were added. In the third phase, the pool was refilled and the dam was removed. Then the spent fuel was moved into the completed end. In the fourth phase, the dam was replaced and the empty part of the pool was drained. The liner and double-tier rack supports were installed, the pool was refilled, and the dam was removed.The project demonstrated that the modification of existing spent fuel fuel pools for handling double-tier fuel racks is a viable solution for increasing the storage capacity at the reactor

  6. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Pattantyus, P.

    1998-01-01

    The current status of the Canadian Spent Fuel Management is described. This includes wet and dry interim storage, transportation issues and future plans regarding final disposal based on deep underground emplacement in stable granite rock. Extension of wet interim storage facilities is not planned, as dry storage technologies have found wide acceptance. (author)

  7. Characteristics of spent nuclear fuel

    International Nuclear Information System (INIS)

    Notz, K.J.

    1988-04-01

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

  8. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  9. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

    Best, R.E.; Garrison, R.F.

    1978-01-01

    This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

  10. Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

    International Nuclear Information System (INIS)

    Liu, Shichang; Li, Zeguang; Wang, Kan; Cheng, Quan; She, Ding

    2018-01-01

    Highlights: •A new random geometry was developed in RMC for mixed and polytype particle/pebble. •This capability was applied to the full core calculations of HTR-10 benchmark. •Reactivity, temperature coefficient and control rod worth of HTR-10 were compared. •This method can explicitly model different packing fraction of different pebbles. •Monte Carlo code with this method can simulate polytype particle/pebble type reactor. -- Abstract: With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

  11. Aeronautical Information System Replacement -

    Data.gov (United States)

    Department of Transportation — Aeronautical Information System Replacement is a web-enabled, automation means for the collection and distribution of Service B messages, weather information, flight...

  12. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  13. Regulation of HtrA2 on WT1 gene expression under imatinib stimulation and its effects on the cell biology of K562 cells.

    Science.gov (United States)

    Zhang, Lixia; Li, Yan; Li, Xiaoyan; Zhang, Qing; Qiu, Shaowei; Zhang, Qi; Wang, Min; Xing, Haiyan; Rao, Qing; Tian, Zheng; Tang, Kejing; Wang, Jianxiang; Mi, Yingchang

    2017-09-01

    The aim of the present study was to investigate the regulation of Wilms Tumor 1 (WT1) by serine protease high-temperature requirement protein A2 (HtrA2), a member of the Htr family, in K562 cells. In addition, the study aimed to observe the effect of this regulation on cell biological functions and its associated mechanisms. Expression of WT1 and HtrA2 mRNA, and proteins following imatinib and the HtrA2 inhibitor 5-[5-(2-nitrophenyl) furfuryl iodine]-1, 3-diphenyl-2-thiobarbituric acid (UCF-101) treatment was detected with reverse transcription-quantitative polymerase chain reaction and western blot analysis. Subsequent to treatment with drugs and UCF-101, the proliferative function of K562 cells was detected using MTT assays, and the rate of apoptosis was detected using Annexin V with propidium iodide flow cytometry in K562 cells. The protein levels in the signaling pathway were analyzed using western blotting following treatment with imatinib and UCF-101. In K562 cells, imatinib treatment activated HtrA2 gene at a transcription level, while the WT1 gene was simultaneously downregulated. Following HtrA2 inhibitor (UCF-101) treatment, the downregulation of WT1 increased gradually. At the protein level, imatinib induced the increase in HtrA2 protein level and concomitantly downregulated WT1 protein level. Subsequent to HtrA2 inhibition by UCF-101, the WT1 protein level decreased temporarily, but eventually increased. Imatinib induced apoptosis in K562 cells, but this effect was attenuated by the HtrA2 inhibitor UCF-101, resulting in the upregulation of the WT1 protein level. However; UCF-101 did not markedly change the proliferation inhibition caused by imatinib. Imatinib activated the p38 mitogen activated protein kinase (p38 MAPK) signaling pathway in K562 cells, and UCF-101 affected the activation of imatinib in the p38 MAPK signaling pathway. Imatinib inhibited the extracellular signal-related kinase (ERK1/2) pathway markedly and persistently, but UCF-101

  14. Collective processing device for spent fuel

    International Nuclear Information System (INIS)

    Irie, Hiroaki; Taniguchi, Noboru.

    1996-01-01

    The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

  15. Radiation Source Replacement Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, Jeffrey W.; Moran, Traci L.; Bond, Leonard J.

    2010-12-01

    This report summarizes a Radiation Source Replacement Workshop in Houston Texas on October 27-28, 2010, which provided a forum for industry and researchers to exchange information and to discuss the issues relating to replacement of AmBe, and potentially other isotope sources used in well logging.

  16. On the correctness of the thermoluminescent high-temperature ratio (HTR) method for estimating ionization density effects in mixed radiation fields

    International Nuclear Information System (INIS)

    Bilski, Pawel

    2010-01-01

    The high-temperature ratio (HTR) method which exploits changes in the LiF:Mg,Ti glow-curve due to high-LET radiation, has been used for several years to estimate LET in an unknown radiation field. As TL efficiency is known to decrease after doses of densely ionizing radiation, a LET estimate is used to correct the TLD-measured values of dose. The HTR method is purely empirical and its general correctness is questionable. The validity of the HTR method was investigated by theoretical simulation of various mixed radiation fields. The LET eff values estimated with the HTR method for mixed radiation fields were found in general to be incorrect, in some cases underestimating the true values of dose-averaged LET by an order of magnitude. The method produced correct estimates of average LET only in cases of almost mono-energetic fields (i.e. in non-mixed radiation conditions). The value of LET eff found by the HTR method may therefore be treated as a qualitative indicator of increased LET, but not as a quantitative estimator of average LET. However, HTR-based correction of the TLD-measured dose value (HTR-B method) was found to be quite reliable. In all cases studied, application of this technique improved the result. Most of the measured doses fell within 10% of the true values. A further empirical improvement to the method is proposed. One may therefore recommend the HTR-B method to correct for decreased TL efficiency in mixed high-LET fields.

  17. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  18. Spent fuel receipt scenarios study

    International Nuclear Information System (INIS)

    Ballou, L.B.; Montan, D.N.; Revelli, M.A.

    1990-09-01

    This study reports on the results of an assignment from the DOE Office of Civilian Radioactive Waste Management to evaluate of the effects of different scenarios for receipt of spent fuel on the potential performance of the waste packages in the proposed Yucca Mountain high-level waste repository. The initial evaluations were performed and an interim letter report was prepared during the fall of 1988. Subsequently, the scope of work was expanded and additional analyses were conducted in 1989. This report combines the results of the two phases of the activity. This study is a part of a broader effort to investigate the options available to the DOE and the nuclear utilities for selection of spent fuel for acceptance into the Federal Waste Management System for disposal. Each major element of the system has evaluated the effects of various options on its own operations, with the objective of providing the basis for performing system-wide trade-offs and determining an optimum acceptance scenario. Therefore, this study considers different scenarios for receipt of spent fuel by the repository only from the narrow perspective of their effect on the very-near-field temperatures in the repository following permanent closure. This report is organized into three main sections. The balance of this section is devoted to a statement of the study objective, a summary of the assumptions. The second section of the report contains a discussion of the major elements of the study. The third section summarizes the results of the study and draws some conclusions from them. The appendices include copies of the waste acceptance schedule and the existing and projected spent fuel inventory that were used in the study. 10 refs., 27 figs

  19. Extended storage of spent fuel

    International Nuclear Information System (INIS)

    1992-10-01

    This document is the final report on the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel and Storage Facility Components during Long Term Storage (BEFAST-II, 1986-1991). It contains the results on wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries who participated in the co-ordinated research programme. Considerable quantities of spent fuel continue to arise and accumulate. Many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology with the construction of additional away-from-reactor storage pools. However, dry storage is increasingly used with most participants considering dry storage concepts for the longer term. Depending on the cladding type options of dry storage in air or inert gas are proposed. Dry storage is becoming widely used as a supplement to wet storage for zirconium alloy clad oxide fuels. Storage periods as long as under wet conditions appear to be feasible. Dry storage will also continue to be used for Al clad and Magnox type fuel. Enhancement of wet storage capacity will remain an important activity. Rod consolidation to increase wet storage capacity will continue in the UK and is being evaluated for LWR fuel in the USA, and may start in some other countries. High density storage racks have been successfully introduced in many existing pools and are planned for future facilities. For extremely long wet storage (≥50 years), there is a need to continue work on fuel integrity investigations and LWR fuel performance modelling. it might be that pool component performance in some cases could be more limiting than the FA storage performance. It is desirable to make concerted efforts in the field of corrosion monitoring and prediction of fuel cladding and poll component behaviour in order to maintain good experience of wet storage. Refs, figs and tabs

  20. Seismic analysis, support design and stress calculation of HTR-PM transport and conversion devices

    International Nuclear Information System (INIS)

    Zhang Zheyu; Yuan Chaolong; Zhang Haiquan; Nie Junfeng

    2012-01-01

    Background: The transport and conversion devices are important guarantees for normal operation of HTR-PM fuel handling system in normal and fault conditions. Purpose: A conflict of devices' support design needs to be solved. The flexibility of supports is required because of pipe thermal expansion displacement, while the stiffness is also required because of large devices quality and eccentric distance. Methods: In this paper, the numerical simulation was employed to analyze the seismic characteristics and optimize the support program, Under the chosen support program, the stress calculation of platen support bracket was designed by solidworks software. Results: The supports solved the conflict between the flexibility and stiffness requirements. Conclusions: Therefore, it can ensure the safety of transport and conversion devices and the supports in seismic conditions. (authors)

  1. The influence of wall materials on Cs depositions in HTR coolant loops

    International Nuclear Information System (INIS)

    Herion, J.

    1975-01-01

    The basic concepts on the effect of the wall material on the deposition of fission products in high temperature reactor (HTR) coolant loops are developed which include the mechanisms of adsorption, solubility and diffusion. General mathematical interrelations of the Cs-adsorption on technical metals are presented and discussed using experimental data from the literature. Desorption energies and frequency factors are determined from measurements of the electrons' work function of metals in Cs atmosphere from R.G. Wilson using these mathematical interrelations. The solubilities and the diffusion constants of Cs are so small in the few investigated cases (Mo, Ta) that high diffusivity paths must be taken into account. Thermochemical considerations on the influence of gaseous impurities on the deposition behaviour are lacking in reliable data. (orig./LH) [de

  2. Concept licensing procedure for an HTR-module nuclear power plant

    International Nuclear Information System (INIS)

    Brinkmann, G.; Will, M.

    1990-01-01

    In April 1987 the companies Siemens and Interatom applied in the West German state of Lower Saxony for a concept licensing procedure to be initiated for an HTR-Module nuclear power plant. In addition to a safety analysis report, numerous additional papers were submitted to the authorized experts. In April 1989 proceedings were suspended for political and legal reasons. By this time both the fire protection report and the plant security concept report had been completed. The safety concept review was continued by order of the Federal Minister for Research and Technology. The draft safety concept report was completed in July 1989. The final version was completed at the end of 1989. (orig.)

  3. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  4. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  5. Actual characteristics study on HTR-10GT coupling with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Peng Xuechuang; Zhu Shutang; Wang Jie

    2005-01-01

    Compared with a plant of steam turbine cycle, a HTGR plant with direct gas turbine cycle has a higher thermal efficiency. A lot of investigations on the characteristics of HTR-10GT, which is the reactor studying project of Tsinghua University, have been carried out, however, all of them are based on the theoretical Brayton Cycle which neglects many actual conditions, such as leakage, pressure loss and so on. For engineering practices, leakage is an unavoidable problem. The difference of the location and capacity of leakage will directly influence the working medium's thermoparameters and lead to fall of the cycle efficiency. The present study is focused on the performance of an actual Brayton cycle with practical conditions of leakage. The present study which based on building the physical and mathematical model of the leakage, aims to study the actual characteristics of the direct gas turbine circle. (authors)

  6. Study on "1"4C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  7. Management system and potential markets for a HTR-GT plant

    International Nuclear Information System (INIS)

    Crommelin, G.A.K.

    1997-01-01

    This article will discuss some aspects which could be helpful to execute a HTR-GT study successfully: 1. The preferred type of organisation for such a study; in order to achieve a maximum of support in society and industry, a minimum of through life costing and a maximum of through life support. 2. The lead time needed for such studies i.e. the design, component testing, prototype testing, the required efficiency, the type of energy in quantity and quality, financial targets, controllability, maintainability and reliability. 3. The potential markets for the nuclear gasturbine driven energy plants in the low power range. Analyses of the markets will be explained from the user's point of view on why, when and how, for what purpose, in which power range, as well as how many units per application would be required. (author)

  8. Status of IAEA international data base on irradiated graphite properties with respect to HTR engineering issues

    International Nuclear Information System (INIS)

    Hacker, P.J.; Haag, G.

    2002-01-01

    The International Database on Irradiated Nuclear Graphite Properties contains data on the physical, chemical, mechanical and other relevant properties of graphites. Its purpose is to provide a platform that makes these properties accessible to approved users in the fields of nuclear power, nuclear safety and other nuclear science and technology applications. The database is constructed using Microsoft Access 97 software and has a controlled distribution by CD ROM to approved users. This paper describes the organisation and management of the database through administrative arrangements approved by the IAEA. It also outlines the operation of the database. The paper concludes with some remarks upon and illustrations of the usefulness of the database for the design and operation of HTR. (authors)

  9. The use and development of the high-temperature reactor (HTR) in China. A conference report

    International Nuclear Information System (INIS)

    Marnet, C.

    2001-01-01

    Gas-cooled graphite-moderated reactors have been under development since the early days of nuclear technology. Starting with plants in Britain and France, reactors employing this combination of coolant and moderator were used in commercial nuclear power plants in the second half of the fifties. At the same time, efforts seeking to use inert helium gas as a coolant resulted in the construction in several countries, the United States and Germany in particular, of larger nuclear power plants with higher coolant temperatures and the resultant thermodynamic advantages of high efficiencies and the option of process heat generation. Economic and political considerations led to the decommissioning of these plants. Today, research and development of high-temperature reactors are concentrated on smaller units. Work is carried out in close international cooperation, especially on project designs and newly commissioned plants in China (HTR-10), Japan (HTTR, Oarai), And South Africa (ESKOM project), but also in the USA and in Russia. (orig.) [de

  10. Reuse of Hydrotreating Spent Catalyst

    International Nuclear Information System (INIS)

    Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

    2004-01-01

    All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

  11. Spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Peev, P.; Kalimanov, N.

    1999-01-01

    The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)

  12. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

    O'Toole, T.

    1993-11-01

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary's initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group's Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities

  13. Spent nuclear fuel storage vessel

    International Nuclear Information System (INIS)

    Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

    1997-01-01

    Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)

  14. Modular dry storage of spent fuel

    International Nuclear Information System (INIS)

    Baxter, J.W.

    1982-01-01

    Long term uncertainties in US spent fuel reprocessing and storage policies and programs are forcing the electric utilities to consider means of storing spent fuel at the reactor site in increasing quantitities and for protracted periods. Utilities have taken initial steps in increasing storage capacity. Existing wet storage pools have in many cases been reracked to optimize their capacity for storing spent fuel assemblies

  15. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

    1990-02-01

    Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

  16. Mechanical Properties and Structures of Pyrolytic Carbon Coating Layer in HTR Coated Particle Fuel

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Young Min; Kim, Woong Ki; Cho, Moon Sung

    2009-01-01

    The TRISO(tri-isotropic)-coated fuel particle for a HTR(High Temperature gas-cooled Reactor) has a diameter of about 1 mm, composed of a nuclear fuel kernel and four different outer coating layers, consisting of a buffer PyC (pyrolytic carbon) layer, inner PyC layer, SiC layer, and outer PyC layer with different coating thicknesses following a specific fuel design. While the fuel kernel is a source for a heat generation by a nuclear fission of fissile uranium, each of the four coating layers acts as a different role in view of retaining the generated fission products and the other interactions during an in-reactor service. Among these coating layers, PyC properties are scarcely in agreement among various investigators and the dependency of their changes upon the deposition condition is comparatively large due to their additional anisotropic properties. Although a recent review work has contributed to an establishment of relationship between the material properties and QC measurements, the data on the mechanical properties and structural parameters of PyC coating layers remain still unclearly evaluated. A review work on dimensional changes of PyC by neutron irradiation was one of re-evaluative works recently attempted by the authors. In this work, an attempt was made to analyze and re-evaluate the existing data of the experimental results of the mechanical properties, i.e., Young's modulus and fracture stress, in relation with the coating conditions, density and the BAF (Bacon Anisotropy Factor), an important structural parameter, of PyC coating layers obtained from various experiments performed in the early periods of the HTR coated particle development

  17. Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

    2015-11-15

    Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

  18. Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

    Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

  19. Euratom research and training in generation IV systems with emphasis on V/HTR

    International Nuclear Information System (INIS)

    Goethem, G. van; Manolatos, P.; Fuetterer, M.

    2006-01-01

    In this overview paper, the following questions are addressed: (1) What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2) What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle: Education, Research, and Innovation') respond to the S/T challenges: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is principally under the responsibility of the 2 Directorates Generals (DG) DG Research (RTD, located in Brussels), which implements and manages the programme of 'indirect actions', and the DG Joint Research Centre (JRC, headquarters in Brussels and 7 scientific institutes in 5 Member States) which carries out 'direct actions' in their own laboratories. In this HTR-2006 introductory paper, the emphasis is on the indirect and direct actions of the 6 th Euratom research framework programme 2003-2006, FP-6, with special emphasis on V/HTR Generation IV research. (orig.)

  20. Safety study for HTR conceptual designs under German siting conditions. Phase I B, specialized volume I

    International Nuclear Information System (INIS)

    1982-08-01

    The basic methodology for determining sequences of events and their frequencies (events and fault trees) does not differ significantly from that of other risk studies. This applies analogously to the treatment of statistical data uncertainties and the description of results in the form of expected value with uncertainty factor. System unavailabilities are determined by means of failure rates, most of which originate from the German Risk Study, and consecutive test intervals. Unlike in other risk studies, common mode failures of components of the same kind are being considered by a mostly 10% fraction of the overall failure of the multi-train system (β-factor). A multitude of planned or unplanned operator actions are identified in the study. They are assessed using models from AIPA and according to WASH-1400. HTR-specific aspects allow mitigating operator actions in the range of days, which are approximately covered by subjective estimates, and extensive reversibility of human errors. British experience with gas-cooled reactors proved to be useful for HTR-specific components. Rates of 0.2 to 1 for small leaks and 1.5 x 10 -3 per reactor-year for larger leaks (tube ruptures) are derived on the basis of 2000 steam generator operating years. Failures of the main blowers (0.1 per blower-year) are covered by other transient events. The behaviour of structural components is of great significance for the progression of core heatup accidents. The liner of the reactor pressure vessel and the concrete located behind will fail over a large area due to decreasing strength at temperatures above 800 0 C. A rupture of closure plugs may be virtually precluded. This also applies to a failure of the reactor containment at internal design pressure. The ultimate strength will only be reached at pressures of more than 14 bar. (orig.) [de

  1. Prediction of extracellular proteases of the human pathogen Helicobacter pylori reveals proteolytic activity of the Hp1018/19 protein HtrA.

    Directory of Open Access Journals (Sweden)

    Martin Löwer

    Full Text Available Exported proteases of Helicobacter pylori (H. pylori are potentially involved in pathogen-associated disorders leading to gastric inflammation and neoplasia. By comprehensive sequence screening of the H. pylori proteome for predicted secreted proteases, we retrieved several candidate genes. We detected caseinolytic activities of several such proteases, which are released independently from the H. pylori type IV secretion system encoded by the cag pathogenicity island (cagPAI. Among these, we found the predicted serine protease HtrA (Hp1019, which was previously identified in the bacterial secretome of H. pylori. Importantly, we further found that the H. pylori genes hp1018 and hp1019 represent a single gene likely coding for an exported protein. Here, we directly verified proteolytic activity of HtrA in vitro and identified the HtrA protease in zymograms by mass spectrometry. Overexpressed and purified HtrA exhibited pronounced proteolytic activity, which is inactivated after mutation of Ser205 to alanine in the predicted active center of HtrA. These data demonstrate that H. pylori secretes HtrA as an active protease, which might represent a novel candidate target for therapeutic intervention strategies.

  2. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel storage rack

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mohd Fazli Zakaria; Zaredah Hashim; Ahmad Nabil Ab Rahim; Mohammad Suhaimi Kassim

    2010-01-01

    PUSPATI TRIGA Reactor (RTP) is a pool type research reactor with 1MW thermal power. It has been safely operated since 28 June 1982. During 28 years of safe operation, there are several systems and components of the RTP that have been maintained, repaired, upgraded and replaced in order to maintain its function and safety conditions. RTP has been proposed to be upgraded so that optimum operation of RTP could be achieved as well as fulfill the future needs. Thus, competencies and technical capabilities were needed to design and develop the reactor system. In the meantime, there is system or component need to be maintained such as fuel elements. Since early operation, most of the fuel elements still can be used and none of the fuel elements was replaced or sent for reprocessing and final disposal. Towards the power upgrading, preparation of spent fuel storage is needed for temporary storing of the fuels discharged from the reactor core. The spent fuel storage rack will be located in the spent fuel pool to accommodate the spent fuels before it is send to reprocessing or final disposal. This paper proposes the conceptual design of the spent fuel storage rack. The output of this paper focused on the physical and engineering design of the spent fuel storage. (author)

  3. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  4. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G.

    2007-06-01

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm 2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  5. LWR Spent Fuel Management for the Smooth Deployment of FBR

    International Nuclear Information System (INIS)

    Fukasawa, T.; Yamashita, J.; Hoshino, K.; Sasahira, A.; Inoue, T.; Minato, K.; Sato, S.

    2015-01-01

    Fast breeder reactors (FBR) and FBR fuel cycle are indispensable to prevent the global warming and to secure the long-term energy supply. Commercial FBR expects to be deployed from around 2050 until around 2110 in Japan by the replacement of light water reactors (LWR) after their 60 years life. The FBR deployment needs Pu (MOX) from the LWR-spent fuel (SF) reprocessing. As Japan can posses little excess Pu, its balance control is necessary between LWR-SF management (reprocessing) and FBR deployment. The fuel cycle systems were investigated for the smooth FBR deployment and the effectiveness of proposed flexible system was clarified in this work. (author)

  6. Measuring the energy spent by parturient women in fasting and in ingesting caloric replacement (HONEY Mensuración de la energia despendida en el ayuno y en el aporte calórico (MIEL en parturientas Mensuração da energia despendida no jejum e no aporte calórico (MEL em parturientes

    Directory of Open Access Journals (Sweden)

    Célia Regina Maganha e Melo

    2007-08-01

    Full Text Available This research aims to measure the energy spending in parturient women of low gestation risk. Participants were selected randomly and submitted to fasting (n=15; Group I or honey ingestion (n=15; Group II. Data were collected by means of capillary blood values and heart frequency monitoring. The paired t-test with a 5% significance level and Tukey's method were used in statistical analysis. The results showed that honey ingestion did not promote an overload in the mother's glucose; the lactate response demonstrated that the substrate offered was well used; the cardiorespiratory rate demonstrated good performance for both groups; the total energy spent during labor demonstrated that carbohydrate ingestion exerts significant influence, improving maternal anaerobic performance; the group which remained in fasting presented, immediately after labor, higher levels of lactate, showing the organism's efforts to compensate for the energy spent.Este estudio tiene como objetivo mensurar el gasto energético de parturientas de bajo riesgo de gestación. Las participantes han sido seleccionadas en dos grupos de manera aleatoria y sometidas a ayuno (n=15; Grupo I e ingestión de miel (n=15; Grupo II. Los datos han sido colectados a partir de los valores de la sangre capilar y monitor de frecuencia cardíaca. Para el análisis estadística han sido empleados el test t pareado, y el método de Tukey. Los resultados han mostrado que la ingestión de miel no provocó sobrecarga en la glicemia materna; la respuesta del lactato demostró que el substrato ofrecido fue bien utilizado; los índices de capacidad cardiorrespiratoria han demostrado buen desempeño para los dos grupos; el gasto energético total durante el trabajo de parto demuestra que la ingestión de carbohidrato tiene influencia significativa, mejorando el desempeño anaeróbico materno; el grupo que ha permanecido en ayuno presentó, inmediatamente después del parto, niveles de lactato más altos

  7. Could wind replace nuclear?

    International Nuclear Information System (INIS)

    2017-01-01

    This article aims at assessing the situation produced by a total replacement of nuclear energy by wind energy, while facing consumption demand at any moment, notably in December. The authors indicate the evolution of the French energy mix during December 2016, and the evolution of the rate between wind energy production and the sum of nuclear and wind energy production during the same month, and then give briefly some elements regarding necessary investments in wind energy to wholly replace nuclear energy. According to them, such a replacement would be ruinous

  8. Characterization program management plan for Hanford K basin spent nuclear fuel

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    1999-01-01

    The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng

  9. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

    Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

    1979-01-01

    Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

  10. Overview of spent fuel management and problems

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Ernst, P.C.

    1998-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel worldwide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. Some projections of spent fuel inventories to the year 2006 are presented and discussed. (author)

  11. Surveillance instrumentation for spent-fuel safeguards

    International Nuclear Information System (INIS)

    McKenzie, J.M.; Holmes, J.P.; Gillman, L.K.; Schmitz, J.A.; McDaniel, P.J.

    1978-01-01

    The movement, in a facility, of spent reactor fuel may be tracked using simple instrumentation together with a real time unfolding algorithm. Experimental measurements, from multiple radiation monitors and crane weight and position monitors, were obtained during spent fuel movements at the G.E. Morris Spent-Fuel Storage Facility. These data and a preliminary version of an unfolding algorithm were used to estimate the position of the centroid and the magnitude of the spent fuel radiation source. Spatial location was estimated to +-1.5 m and source magnitude to +-10% of their true values. Application of this surveillance instrumentation to spent-fuel safeguards is discussed

  12. Association of serotonin transporter (SLC6A4 & receptor (5HTR1A, 5HTR2A polymorphisms with response to treatment with escitalopram in patients with major depressive disorder : A preliminary study

    Directory of Open Access Journals (Sweden)

    Aniruddha Basu

    2015-01-01

    Full Text Available Background & objectives: Genetic factors have potential of predicting response to antidepressants in patients with major depressive disorder (MDD. In this study, an attempt was made to find an association between response to escitalopram in patients with MDD, and serotonin transporter (SLC6A4 and receptor (5HTR1A, 5HTR2A polymorphisms. Methods: Fifty five patients diagnosed as suffering from MDD, were selected for the study. The patients were treated with escitalopram over a period of 6-8 wk. Severity of depression, response to treatment and side effects were assessed using standardised instruments. Genetic variations from HTR1A (rs6295, HTR2A (rs6311 and rs6313 and SLC6A4 (44 base-pair insertion/deletion at 5-HTTLPR were genotyped. The genetic data of the responders and non-responders were compared to assess the role of genetic variants in therapeutic outcome. Results: Thirty six (65.5% patients responded to treatment, and 19 (34.5% had complete remission. No association was observed for genotype and allelic frequencies of single nucleotide polymorphisms (SNPs among remitter/non-remitter and responder/non-responder groups, and six most common side-effects, except memory loss which was significantly associated with rs6311 ( p0 =0.03. Interpretation & conclusions: No significant association was found between the SNPs analysed and response to escitalopram in patients with MDD though a significant association was seen between the side effect of memory loss and rs6311. Studies with larger sample are required to find out genetic basis of antidepressant response in Indian patients.

  13. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

  14. A Monte Carlo based spent fuel analysis safeguards strategy assessment

    International Nuclear Information System (INIS)

    Fensin, Michael L.; Tobin, Stephen J.; Swinhoe, Martyn T.; Menlove, Howard O.; Sandoval, Nathan P.

    2009-01-01

    Safeguarding nuclear material involves the detection of diversions of significant quantities of nuclear materials, and the deterrence of such diversions by the risk of early detection. There are a variety of motivations for quantifying plutonium in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguards nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from spent fuel; however, no single NDA technique can, in isolation, quantify elemental plutonium and other actinides of interest in spent fuel. A study has been undertaken to determine the best integrated combination of cost effective techniques for quantifying plutonium mass in spent fuel for nuclear safeguards. A standardized assessment process was developed to compare the effective merits and faults of 12 different detection techniques in order to integrate a few techniques and to down-select among the techniques in preparation for experiments. The process involves generating a basis burnup/enrichment/cooling time dependent spent fuel assembly library, creating diversion scenarios, developing detector models and quantifying the capability of each NDA technique. Because hundreds of input and output files must be managed in the couplings of data transitions for the different facets of the assessment process, a graphical user interface (GUI) was development that automates the process. This GUI allows users to visually create diversion scenarios with varied replacement materials, and generate a MCNPX fixed source detector assessment input file. The end result of the assembly library assessment is to select a set of common source terms and diversion scenarios for quantifying the capability of each of the 12 NDA techniques. We present here the generalized

  15. A Monte Carlo Based Spent Fuel Analysis Safeguards Strategy Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Fensin, Michael L.; Tobin, Stephen J.; Swinhoe, Martyn T.; Menlove, Howard O.; Sandoval, Nathan P. [Los Alamos National Laboratory, E540, Los Alamos, NM 87545 (United States)

    2009-06-15

    Safeguarding nuclear material involves the detection of diversions of significant quantities of nuclear materials, and the deterrence of such diversions by the risk of early detection. There are a variety of motivations for quantifying plutonium in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguards nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from spent fuel; however, no single NDA technique can, in isolation, quantify elemental plutonium and other actinides of interest in spent fuel. A study has been undertaken to determine the best integrated combination of cost effective techniques for characterizing Pu mass in spent fuel for nuclear safeguards. A standardized assessment process was developed to compare the effective merits and faults of 12 different detection techniques in order to integrate a few techniques and to down-select among the techniques in preparation for experiments. The process involves generating a basis burnup/enrichment/cooling time dependent spent fuel assembly library, determining and identifying limiting diversion scenarios, developing detector models and quantifying the capability of each NDA technique. Because hundreds of input and output files must be managed in the couplings of data transitions for the different facets of the assessment process, a graphical user interface (GUI) was development that automates the process. This GUI allows users to visually create diversion scenarios with varied replacement materials, and generate a MCNPX fixed source detector assessment input file. The end result of the assembly library assessment is to select a set of common source terms and diversion scenarios for quantifying the capability of each of the 12 NDA techniques. We present here

  16. Slab replacement maturity guidelines.

    Science.gov (United States)

    2014-04-01

    This study investigated the use of maturity method to determine early age strength of concrete in slab : replacement application. Specific objectives were (1) to evaluate effects of various factors on the compressive : maturity-strength relationship ...

  17. Partial knee replacement

    Science.gov (United States)

    ... good range of motion in your knee. The ligaments in your knee are stable. However, most people with knee arthritis have a surgery called a total knee arthroplasty (TKA). Knee replacement is most often done in people age 60 ...

  18. Carbohydrates as Fat Replacers.

    Science.gov (United States)

    Peng, Xingyun; Yao, Yuan

    2017-02-28

    The overconsumption of dietary fat contributes to various chronic diseases, which encourages attempts to develop and consume low-fat foods. Simple fat reduction causes quality losses that impede the acceptance of foods. Fat replacers are utilized to minimize the quality deterioration after fat reduction or removal to achieve low-calorie, low-fat claims. In this review, the forms of fats and their functions in contributing to food textural and sensory qualities are discussed in various food systems. The connections between fat reduction and quality loss are described in order to clarify the rationales of fat replacement. Carbohydrate fat replacers usually have low calorie density and provide gelling, thickening, stabilizing, and other texture-modifying properties. In this review, carbohydrates, including starches, maltodextrins, polydextrose, gums, and fibers, are discussed with regard to their interactions with other components in foods as well as their performances as fat replacers in various systems.

  19. Hip joint replacement - slideshow

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/presentations/100006.htm Hip joint replacement - series—Normal anatomy To use the ... to slide 5 out of 5 Overview The hip joint is made up of two major parts: ...

  20. Tool Inventory and Replacement

    Science.gov (United States)

    Bear, W. Forrest

    1976-01-01

    Vocational agriculture teachers are encouraged to evaluate curriculum offerings, the new trends in business and industry, and develop a master tool purchase and replacement plan over a 3- to 5-year period. (HD)

  1. Knee joint replacement

    Science.gov (United States)

    ... to make everyday tasks easier. Practice using a cane, walker , crutches , or a wheelchair correctly. On the ... ask your doctor Knee joint replacement - discharge Preventing falls Preventing falls - what to ask your doctor Surgical ...

  2. Product Platform Replacements

    DEFF Research Database (Denmark)

    Sköld, Martin; Karlsson, Christer

    2012-01-01

    . To shed light on this unexplored and growing managerial concern, the purpose of this explorative study is to identify operational challenges to management when product platforms are replaced. Design/methodology/approach – The study uses a longitudinal field-study approach. Two companies, Gamma and Omega...... replacement was chosen in each company. Findings – The study shows that platform replacements primarily challenge managers' existing knowledge about platform architectures. A distinction can be made between “width” and “height” in platform replacements, and it is crucial that managers observe this in order...... to challenge their existing knowledge about platform architectures. Issues on technologies, architectures, components and processes as well as on segments, applications and functions are identified. Practical implications – Practical implications are summarized and discussed in relation to a framework...

  3. Flood control construction of Shidao Bay nuclear power plant and safety analysis for hypothetical accident of HTR-PM

    International Nuclear Information System (INIS)

    Chen Yongrong; Zhang Keke; Zhu Li

    2014-01-01

    A series of events triggered by tsunami eventually led to the Fukushima nuclear accident. For drawing lessons from the nuclear accident and applying to Shidao Bay nuclear power plant flood control construction, we compare with the state laws and regulations, and prove the design of Shidao Bay nuclear power plant flood construction. Through introducing the history of domestic tsunamis and the national researches before and after the Fukushima nuclear accident, we expound the tsunami hazards of Shidao Bay nuclear power plant. In addition, in order to verify the safety of HTR-PM, we anticipate the contingent accidents after ''superposition event of earthquake and extreme flood'', and analyse the abilities and measures of HTR-PM to deal with these beyond design basis accidents (BDBA). (author)

  4. Application of grey model on analyzing the passive natural circulation residual heat removal system of HTR-10

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; PENG Changhong; WANG Zenghui; WANG Ruosu

    2008-01-01

    Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR (High Temperature gas-cooled Reactor),the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM(1,1) and GM(1, 3) model are built based on the inlet air temperature of chimney, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an important role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR.

  5. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  6. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  7. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  8. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

    Bergmann, D.W.

    1995-01-01

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation

  9. Dry storage of spent fuel

    International Nuclear Information System (INIS)

    Jeffrey, R.

    1993-01-01

    Scottish Nuclear's plans to build and operate dry storage facilities at each of its two nuclear power station sites in Scotland are explained. An outline of where waste materials arise as part of the operation and decommissioning of nuclear power stations, the volumes for each category of high-, intermediate-and low-level wastes and the costs involved are given. The present procedure for the spent fuels from Hunterston-B and Torness stations is described and Scottish Nuclear's aims of driving output up and costs down are studied. (UK)

  10. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  11. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

    1987-09-01

    Historical inventories of spent fuel and utility estimates of future discharges from US commercial nuclear reactors are presented through the year 2005. The ultimate needs for additional storage capacity are estimated. These estimtes are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December, 1986, and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to the DOE Energy Information Administration (EIA) through the 1987 RW-859 data survey. 14 refs., 4 figs., 9 tabs

  12. CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

    International Nuclear Information System (INIS)

    Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe; Perez, Marc; Cellier, Francois; Vitali, Marie-Pierre

    2006-01-01

    In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO 2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO 2 kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

  13. Measurement of fission gases released by HTR irradiation samples in the BR 2 reactor (Mol/Belgium)

    International Nuclear Information System (INIS)

    Koetter, H.; Mueller, H.B.

    1975-04-01

    During the irradiation of HTR-test fuel (coated particles resp. compacts) the release rate of fission gases is measured automatically. For that purpose the γ-spectrum of the sweep gas is analyzed with respect to the isotopes Xe-133, Xe-135, Xe-135sup(m), Xe-137, Xe-138, Kr-85sup(m), Kr-87, Kr-88, Kr-89. Gas sampling, spectral analysis, data handling and evaluation are controlled by computer. (orig.) [de

  14. Studi Awal Desain Pebble Bed Reactor Berbasis Htr-pm Dengan Skema Resirkulasi Bahan Bakar Once-through-then-out

    OpenAIRE

    Setiadipura, Topan; Pane, Jupiter Sitorus; Zuhair, Zuhair

    2016-01-01

    STUDI AWAL DESAIN PEBBLE BED REACTOR BERBASIS HTR-PM DENGAN RESIRKULASI BAHAN BAKAR ONCE-THROUGH-THEN-OUT. Reaktor nuklir tipe pebble bed reactor (PBR) adalah salah satu reaktor canggih dengan fitur keselamatan pasif yang kuat. Pada desain tipe ini berpotensi untuk dilakukan kogenerasi yang bermanfaat untuk pengolahan berbagai mineral di berbagai pulau di Indonesia. Operasi PBR dapat lebih disederhanakan dengan menerapkan skema pengisian bahan bakar once-through-then-out (OTTO) dimana bahan b...

  15. Rapid blockade of telomerase activity and tumor cell growth by the DPL lipofection of ribbon antisense to hTR.

    Science.gov (United States)

    Bajpai, Arun K; Park, Jeong-Hoh; Moon, Ik-Jae; Kang, Hyungu; Lee, Yun-Han; Doh, Kyung-Oh; Suh, Seong-Il; Chang, Byeong-Churl; Park, Jong-Gu

    2005-09-29

    Ribbon antisense (RiAS) to the hTR RNA, a component of the telomerase complex, was employed to inhibit telomerase activity and cancer cell growth. The antisense molecule, hTR-RiAS, combined with enhanced cellular uptake was shown to effectively inhibit telomerase activity and cause rapid cell death in various cancer cell lines. When cancer cells were treated with hTR-RiAS, the level of hTR RNA was reduced by more than 90% accompanied with reduction in telomerase activity. When checked for cancer cell viability, cancer cell lines treated with hTR-RiAS using DNA+Peptide+Lipid complex showed 70-80% growth inhibition in 3 days. The reduced cell viability was due to apoptosis as the percentage of cells exhibiting the sub-G0 arrest and DNA fragmentation increased after antisense treatment. Further, when subcutaneous tumors of a colon cancer cell line (SW480) were treated intratumorally with hTR-RiAS, tumor growth was markedly suppressed with almost total ablation of hTR RNA in the tumor tissue. Cells in the tumor tissue were also found to undergo apoptosis after hTR-RiAS treatment. These results suggest that hTR-RiAS is an effective anticancer reagent, with a potential for broad efficacy to diverse malignant tumors.

  16. Evidence for the effect of serotonin receptor 1A gene (HTR1A) polymorphism on tractability in Thoroughbred horses.

    Science.gov (United States)

    Hori, Y; Tozaki, T; Nambo, Y; Sato, F; Ishimaru, M; Inoue-Murayama, M; Fujita, K

    2016-02-01

    Tractability, or how easily animals can be trained and controlled, is an important behavioural trait for the management and training of domestic animals, but its genetic basis remains unclear. Polymorphisms in the serotonin receptor 1A gene (HTR1A) have been associated with individual variability in anxiety-related traits in several species. In this study, we examined the association between HTR1A polymorphisms and tractability in Thoroughbred horses. We assessed the tractability of 167 one-year-old horses reared at a training centre for racehorses using a questionnaire consisting of 17 items. A principal components analysis of answers contracted the data to five principal component (PC) scores. We genotyped two non-synonymous single nucleotide polymorphisms (SNPs) in the horse HTR1A coding region. We found that one of the two SNPs, c.709G>A, which causes an amino acid change at the intracellular region of the receptor, was significantly associated with scores of four of five PCs in fillies (all Ps Horses carrying an A allele at c.709G>A showed lower tractability. This result provides the first evidence that a polymorphism in a serotonin-related gene may affect tractability in horses with the effect partially different depending on sex. © 2015 Stichting International Foundation for Animal Genetics.

  17. A drying system for spent fuel assemblies

    International Nuclear Information System (INIS)

    Suikki, M.; Warinowski, M.; Nieminen, J.

    2007-06-01

    The report presents a proposed drying apparatus for spent fuel assemblies. The apparatus is used for removing the moisture left in fuel assemblies during intermediate storage and transport. The apparatus shall be installed in connection with the fuel handling cell of an encapsulation plant. The report presents basic requirements for and implementation of the drying system, calculation of the drying process, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components have not been included. The report also describes actions for possible malfunction and fault conditions. An objective of the drying system for fuel assemblies is to remove moisture from the assemblies prior to placing the same in a disposal canister for spent nuclear fuel. Drying is performed as a vacuum drying process for vaporizing and draining the moisture present on the surface of the assemblies. The apparatus comprises two pieces of drying equipment. One of the chambers is equipped to take up Lo1-2 fuel assemblies and the other OL1-2 fuel assemblies. The chambers have an internal space sufficient to accommodate also OL3 fuel assemblies, but this requires replacing the internal chamber structure for laying down the assemblies to be dried. The drying chambers can be closed with hatches facing the fuel handling cell. Water vapour pumped out of the chamber is collected in a controlled manner, first by condensing with a heat exchanger and further by freezing in a cold trap. For reasons of safety, the exhaust air of vacuum pumps is further delivered into the ventilation outlet duct of a controlled area. The adequate drying result is ascertained by a low final pressure of about 100 Pa, as well as by a sufficient holding time. The chamber is built for making its cleaning as easy as possible in the event of a fuel rod breaking during a drying, loading or unloading

  18. Spent nuclear fuel storage - Basic concept

    International Nuclear Information System (INIS)

    Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

    2009-01-01

    According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)

  19. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki.

    1976-01-01

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  20. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Schmitt, D.

    1985-01-01

    How should the decision in favour of reprocessing and against alternative waste management concepts be judged from an economic standpoint. Reprocessing is not imperative neither for resource-economic reasons nor for nuclear energy strategy reasons. On the contrary, the development of an ultimate storage concept representing a real alternative promising to close, within a short period of time, the nuclear fuel cycle at low cost. At least, this is the result of an extensive economic efficiency study recently submitted by the Energy Economics Institute which investigated all waste management concepts relevant for the Federal Republic of Germany in the long run, i.e. direct ultimate storage of spent fuel elements (''Other waste disposal technologies'' - AE) as well as reprocessing of spent fuel elements where re-usable plutonium and uranium are recovered and radioactive waste goes to ultimate storage (''Integrated disposal'' - IE). Despite such fairly evident results, the government of the Federal Republic of Germany has favoured the construction of a reprocessing plant. From an economic point of view there is no final answer to the question whether or not the argumentation is sufficient to justify the decision to construct a reprocessing plant. This is true for both the question of technical feasibility and issues of overriding significance of a political nature. (orig./HSCH) [de

  1. Spent fuel storage criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

    1995-10-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs.

  2. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    Amin, E.M.; Elmessiry, A.M.

    1995-01-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

  3. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Weiner, R.F.

    1993-01-01

    This paper discusses RADTRAN calculational models and parameter values for describing dose to workers during incident-free ship-to-truck transfer of spent fuel. Data obtained during observation of the offloading of research reactor spent fuel at Newport News Terminal in the Port of Hampton Roads, Virginia, are described. These data include estimates of exposure times and distances for handlers, inspectors, and other workers during offloading and overnight storage. Other workers include crane operators, scale operators, security personnel, and truck drivers. The data are compared to the default data in RADTRAN 4, and the latter are found to be conservative. The casks were loaded under IAEA supervision at their point of origin, and three separate radiological inspections of each cask were performed at the entry to the port (Hampton Roads) by the U.S. Coast Guard, the state of Virginia, and the shipping firm. As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handler exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. (author)

  4. HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

    International Nuclear Information System (INIS)

    Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

    2009-01-01

    To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

  5. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

  6. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  7. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  8. Overview on spent fuel management strategies

    International Nuclear Information System (INIS)

    Dyck, P.

    2002-01-01

    This paper presents an overview on spent fuel management strategies which range from reprocessing to interim storage in a centralised facility followed by final disposal in a repository. In either case, more spent fuel storage capacity (wet or dry, at-reactor or away-from-reactor, national or regional) is required as spent fuel is continuously accumulated while most countries prefer to defer their decision to choose between these two strategies. (author)

  9. Spent fuel management in France: Programme status

    International Nuclear Information System (INIS)

    Chaudat, J.P.

    1990-01-01

    France's programme is best characterized as a closed fuel cycle including reprocessing, Plutonium recycling in PWR and use of breeder reactors. The current installed nuclear capacity is 52.5 GWe from 55 units. The spent fuel management scheme chosen is reprocessing. This paper describes the national programme, spent nuclear fuel storage, reprocessing and contracts for reprocessing of spent fuel from various countries. (author). 5 figs, 2 tabs

  10. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  11. The security management of spent filter cartridge in Qinshan phase 3 (heavy water reactor) nuclear power plant

    International Nuclear Information System (INIS)

    Xue Dahai

    2005-01-01

    Qinshan phase 3 nuclear power plant is the first CANDU plant that China fetched in from Canada, and both two units operate under well condition up to now. The radioactive wastes produced during the unit operation mainly include technical waste, spent resin, and spent filter cartridge. The spent filter cartridge is one important part both in the volume and radioactivity of the radioactive waste, and it is the important content of radioactive waste management. Different from PWR, part of high radioactive spent filter in CANDU unit comes from heavy water system such as moderator system. It has to be dried through blowing before replaced from the system. But this working procedure result the filtrate dreg become flexible, and it can bring on the risk of internal or external exposure. It is very important to pay high attention to control the contamination spread during spent filter inside transfer. (authors)

  12. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  13. Spent fuels transportation coming from Australia

    International Nuclear Information System (INIS)

    2002-01-01

    Maritime transportation of spent fuels from Australia to France fits into the contract between COGEMA and ANSTO, signed in 1999. This document proposes nine information cards in this domain: HIFAR a key tool of the nuclear, scientific and technological australian program; a presentation of the ANSTO Australian Nuclear Science and Technology Organization; the HIFAR spent fuel management problem; the COGEMA expertise in favor of the research reactor spent fuel; the spent fuel reprocessing at La Hague; the transports management; the transport safety (2 cards); the regulatory framework of the transports. (A.L.B.)

  14. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  15. Spent fuel critical masses and supportive measurements

    International Nuclear Information System (INIS)

    Toffer, H.; Wells, A.H.

    1987-01-01

    Critical masses for spent fuel are larger than for green fuel and therefore use of the increased masses could result in improved handling, storage, and transport of such materials. To apply spent fuel critical masses requires an assessment of fuel exposure and the corresponding isotopic compositions. The paper discusses several approaches at the Hanford N Reactor in establishing fuel exposure, including a direct measurement of spent to green fuel critical masses. The benefits derived from the use of spent fuel critical masses are illustrated for cask designs at the Nuclear Assurance Corporation. (author)

  16. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  17. Spent fuel's behavior under dynamic drip tests

    International Nuclear Information System (INIS)

    Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

    1995-01-01

    In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

  18. Reprocessing method for spent fuel

    International Nuclear Information System (INIS)

    Fujie, Makoto; Shoji, Yuichi; Kobayashi, Tsuguyuki.

    1997-01-01

    After reducing oxides of uranium (U), plutonium (Pu) and miner actinides in spent fuels by magnesium (Mg) in a molten salt, rear earth element oxides and salts of alkali metals and alkaline earth metals contained in the molten salt phase are separated and removed. Further, the Mg phase containing the reduced metals is evaporated to separate and remove Mg, thereby recovering U, Pu and minor actinides. In a lithium (Li) process, Li 2 O also generated in the reduction step is regenerated to Li simultaneously, and the reduction is conducted while suppressing the Li 2 O concentration in the molten salt low. This can improve the reduction rate of oxides of U, Pu and minor actinides compared with conventional cases. Since Li 2 O is regenerated into Li in the reduction step of the Li process, deposited Li 2 O is not carried to an electrolysis purification step, and recovering rate of U, Pu and minor actinides is not lowered. (T.M.)

  19. Method of decladding spent fuel

    International Nuclear Information System (INIS)

    Fukutome, Kazuyuki; Kitagawa, Kazuo.

    1988-01-01

    Purpose: To enable to safety and easy decladding of nuclear fuels thereby reduce the processing cost. Constitution: Upon dismantling of a spent fuel rod, the fuel rod is heated at least to such a temperature that the ductility of a fuel can is recovered, then transported by using seizing rollers, by which the fuel rod is pressurized from the outer circumference to break the nuclear fuels at the inside thereof. Then, the destructed fuels are recovered from both ends of the fuel can. With such a constitution, since the ductility of the fuel can is recovered by heating, when the fuel rod is passed through the rollers in this state, the fuel can is deformed to destroy the nuclear fuels at the inside thereof. Since the nuclear fuels are destroyed into small pieces, they can be taken out easily from both ends of the fuel can. (Kawakami, Y.)

  20. Assignment of the 5HT7 receptor gene (HTR7) to chromosome 10q and exclusion of genetic linkage with Tourette syndrome

    Energy Technology Data Exchange (ETDEWEB)

    Gelernter, J.; Rao, P.A.; Pauls, D.L. [Yale Univ. School of Medicine, West Haven, CT (United States)] [and others

    1995-03-20

    A novel serotonin receptor designated 5HT7 (genetic locus HTR7) was cloned in 1993. This receptor has interesting properties related to ligand affinity and CNS distribution that render HTR7 a very interesting candidate gene for neuropsychiatric disorders. We mapped this gene, first by physical methods and then by genetic linkage. First, we made a tentative assignment to chromosome 10, based on hybridization of an HTR7 probe to a Southern blot of DNA from somatic cell hybrids. We then identified a genetic polymorphism at the HTR7 locus. We identified one extended pedigree where the polymorphism segregated. Using the LEPED computer program for pairwise linkage analysis, we confirmed the assignment of the gene to chromosome 10, specifically 10q21-q24, based on a lod score of 5.37 at 0% recombination between HTR7 and D10S20 (a chromosome 10 reference marker). Finally, we excluded genetic linkage between this locus and Tourette syndrome under a reasonable set of assumptions. 15 refs., 1 fig., 1 tab.