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Sample records for turbine-modular helium reactor

  1. Gas turbine modular helium reactor in cogeneration

    International Nuclear Information System (INIS)

    Leon de los Santos, G.

    2009-10-01

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  2. Evaluation of the Gas Turbine Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  3. Evaluation of the Gas Turbine Modular Helium Reactor

    International Nuclear Information System (INIS)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs

  4. The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

    International Nuclear Information System (INIS)

    LaBar, Malcolm P.

    2002-01-01

    The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

  5. Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto

    2006-04-15

    Amid generation IV of nuclear power plants, the Gas Turbine - Modular Helium Reactor, designed by General Atomics, is the only core with an energy conversion efficiency of 50%; the safety aspects, coupled to construction and operation costs lower than ordinary Light Water Reactors, renders the Gas Turbine - Modular Helium reactor rather unequaled. In the present studies we investigated the possibility to operate the GT-MHR with two types of fuels: LWRs waste and thorium; since thorium is made of only fertile {sup 232}Th, we tried to mix it with pure {sup 233}U, {sup 235}U or {sup 239}Pu; ex post facto, only uranium isotopes allow the reactor operation, that induced us to examine the possibility to use a mixture of uranium, enriched 20% in {sup 235}U, and thorium. We performed all calculations by the MCNP and MCB codes, which allowed to model the reactor in a very detailed three-dimensional geometry and to describe the nuclides transmutation in a continuous energy approach; finally, we completed our studies by verifying the influence of the major nuclear data libraries, JEFF, JENDL and ENDF/B, on the obtained results.

  6. Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

    2009-10-15

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  7. Presentation summary: Gas Turbine - Modular Helium Reactor (GT-MHR)

    International Nuclear Information System (INIS)

    2001-01-01

    Numerous prototypes and demonstration plants have been constructed and operated beginning with the Dragon plant in the early 1960s. The MHTGR was the U.S. developed modular plant and underwent pre application review by NRC. The GT-MHR represents a further refinement on this concept with the steam cycle being replaced by a closed loop gas turbine (Brayton) cycle. Modular gas reactors and the GT-MHR represent a fundamental shift in reactor design and safety philosophy. The reactor system is contained in a 3 vessel, side-by-side arrangement. The reactor and a shutdown cooling system are in one vessel, and the gas turbine based power conversion system, including the generator, in a second parallel vessel. A more detailed look at the system shows the compact arrangement of gas turbine, compressors, recuperator, heat exchanges, and generator. Fueled blocks are stacked in three concentric rings with inert graphite blocks making up the inner and outer reflectors. Operating control rods are located outside the active core while startup control rods and channels for reserve shutdown pellets are located near the core center. Ceramic coated fuel is the key to the GT-MHR's safety and economics. A kernel of Uranium oxycarbide (or UO 2 ) is placed in a porous carbon buffer and then encapsulated in multiple layers of pyrolytic carbon and silicon carbide. These micro pressure vessels withstand internal pressures of up to 2,000 psi and temperatures of nearly 2,000 C providing extremely resilient containment of fission products under both normal operating and accident conditions. The fuel particles are blended in carbon pitch, forming fuel rods, and then loaded into holes within large graphite fuel elements. Fuel elements are stacked to form the core. Fuel particle testing in has repeatedly demonstrated the high temperature resilience of coated particle fuel to temperature approaching 2,000 C. As an conservative design goal, GT-MHR has been sized to keep maximum fuel temperatures

  8. Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)

    International Nuclear Information System (INIS)

    Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

    1994-04-01

    A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR's safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements

  9. The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

    International Nuclear Information System (INIS)

    Silberstein, A.J.

    1998-01-01

    As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

  10. The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

    International Nuclear Information System (INIS)

    Zgliczynski, J.B.; Silady, F.A.; Neylan, A.J.

    1994-04-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%

  11. Optimization of the gas turbine-modular helium reactor using statistical methods to maximize performance without compromising system design margins

    International Nuclear Information System (INIS)

    Lommers, L.J.; Parme, L.L.; Shenoy, A.S.

    1995-07-01

    This paper describes a statistical approach for determining the impact of system performance and design uncertainties on power plant performance. The objectives of this design approach are to ensure that adequate margin is provided, that excess margin is minimized, and that full advantage can be taken of unconsumed margin. It is applicable to any thermal system in which these factors are important. The method is demonstrated using the Gas Turbine Modular Helium Reactor as an example. The quantitative approach described allows the characterization of plant performance and the specification of the system design requirements necessary to achieve the desired performance with high confidence. Performance variations due to design evolution, inservice degradation, and basic performance uncertainties are considered. The impact of all performance variabilities is combined using Monte Carlo analysis to predict the range of expected operation

  12. Exergoeconomic assessment and parametric study of a Gas Turbine-Modular Helium Reactor combined with two Organic Rankine Cycles

    International Nuclear Information System (INIS)

    Mohammadkhani, F.; Shokati, N.; Mahmoudi, S.M.S.; Yari, M.; Rosen, M.A.

    2014-01-01

    An exergoeconomic analysis is reported for a combined system with a net electrical output of 299 MW in which waste heat from a Gas Turbine-Modular Helium Reactor (GT-MHR) is utilized by two Organic Rankine Cycles (ORCs). A parametric study is also done to reveal the effects on the exergoeconomic performance of the combined system of such significant parameters as compressor pressure ratio, turbine inlet temperature, temperatures of evaporators, pinch point temperature difference in the evaporators and degree of superheat at the ORC (Organic Rankine Cycle) turbines inlet. Finally the combined cycle performance is optimized from the viewpoint of exergoeconomics. The results show that the precooler, the intercooler and the ORC condensers exhibit the worst exergoeconomic performance. For the overall system, the exergoeconomic factor, the capital cost rate and the exergy destruction cost rate are determined to be 37.95%, 6876 $/h and 11,242 $/h, respectively. Also, it is observed that the unit cost of electricity produced by the GT-MHR turbine increases with increasing GT-MHR turbine inlet temperature but decreases as the other above mentioned parameters increase. - Highlights: • An exergoeconomic analysis is performed for the GT-MHR/ORC (Organic Rankine Cycle) combined cycle. • The effects of decision parameters on the exergoeconomic performance are studied. • The highest exergy destructions occur in the precooler, intercooler and condenser. • Superheating the working fluid at the ORC turbine inlet is not necessary. • Thermodynamic and exergoeconomic optimal conditions differ from each other

  13. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167 Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1 B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167 Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff , at the beginning of the operation

  14. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

    2006-01-15

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the {sup 167}Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in {sup 1}B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k {sub eff} value, the {sup 167}Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the {beta} {sub eff}, at the beginning of the operation.

  15. Performance of the gas turbine-modular helium reactor fuelled with different types of fertile TRISO particles

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with 239 Pu, 233 U or 235 U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of 232 Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes ( 239 Pu, 233 U and 235 U), requires a quantity of thorium 25-30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, 233 U, 235 U and 239 Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium

  16. Performance of the gas turbine-modular helium reactor fuelled with different types of fertile TRISO particles

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-106 91 Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-106 91 Stockholm (Sweden)

    2005-11-15

    Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with {sup 239}Pu, {sup 233}U or {sup 235}U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of {sup 232}Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes ({sup 239}Pu, {sup 233}U and {sup 235}U), requires a quantity of thorium 25-30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 {mu}m. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, {sup 233}U, {sup 235}U and {sup 239}Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.

  17. Combining a gas turbine modular helium reactor and an accelerator and for near total destruction of weapons grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, A.M.; Lane, R.K.; Sherman, R. [General Atomics, San Diego, CA (United States)

    1995-10-01

    Fissioning surplus weapons-grade plutonium (WG-Pu) in a reactor is an effective means of rendering this stockpile non-weapons useable. In addition the enormous energy content of the plutonium is released by the fission process and can be captured to produce valuable electric power. While no fission option has been identified that can accomplish the destruction of more than about 70% of the WG-Pu without repeated reprocessing and recycling, which presents additional opportunities for diversion, the gas turbine modular helium-cooled reactor (GT-MHR), using an annular graphite core and graphite inner and outer reflectors combines the maximum plutonium destruction and highest electrical production efficiency and economics in an inherently safe system. Accelerator driven sub-critical assemblies have also been proposed for WG-Pu destruction. These systems offer almost complete WG-Pu destruction, but achieve this goal by using circulating aqueous or molten salt solutions of the fuel, with potential safety implications. By combining the GT-MHR with an accelerator-driven sub-critical MHR assembly, the best features of both systems can be merged to achieve the near total destruction of WG-Pu in an inherently safe, diversion-proof system in which the discharged fuel elements are suitable for long term high level waste storage without the need for further processing. More than 90% total plutonium destruction, and more than 99.9% Pu-239 destruction, could be achieved. The modular concept minimizes the size of each unit so that both the GT-MHR and the accelerator would be straightforward extensions of current technology.

  18. An exergoeconomic assessment of waste heat recovery from a Gas Turbine-Modular Helium Reactor using two transcritical CO_2 cycles

    International Nuclear Information System (INIS)

    Wang, Xurong; Dai, Yiping

    2016-01-01

    Highlights: • A Gas Turbine-Modular Helium Reactor is coupled with two transcritical CO_2 cycles. • Exergoeconomic analysis and optimization of the combined cycle was performed. • The energy efficiency of the cogeneration system was enhanced by 7.92% at 850 °C. • The overall exergoeconomic factor of the system could up to 55.2%. - Abstract: A comprehensive study is performed on an energy conversion system which combines a gas turbine-modular helium reactor (GT-MHR) and two transcritical CO_2 cycles (tCO_2). The aim of this study is to assess the energy, exergy and economic behavior of the proposed system, considering five indicators: the energy efficiency, the exergy efficiency, the total exergy destruction cost rate, the overall exergoeconomic factor and the total cost rate. A parametric study is also conducted to evaluate the influence of key decision variables on the GT-MHR/tCO_2 performance. Finally, the combined cycle is optimized to minimize the total cost rate. The results show that the energy efficiency of GT-MHR/tCO_2 cycle is 7.92% higher than that of the simple GT-MHR cycle at 850 °C. The largest exergy destruction rate takes place in the reactor, and after that in the helium turbine and the recuperator. The components in tCO_2 cycles have less exergy destruction. When the optimization is conducted based on the exergoeconomics, the overall exergoeconomic factor, the total cost rate and the total exergy destruction cost rate are 55.2%, 20,752 $/h and 9292 $/h, respectively.

  19. An exergoeconomic investigation of waste heat recovery from the Gas Turbine-Modular Helium Reactor (GT-MHR) employing an ammonia–water power/cooling cycle

    International Nuclear Information System (INIS)

    Zare, V.; Mahmoudi, S.M.S.; Yari, M.

    2013-01-01

    A detailed exergoeconomic analysis is performed for a combined cycle in which the waste heat from the Gas Turbine-Modular Helium Reactor (GT-MHR) is recovered by an ammonia–water power/cooling cogeneration system. Parametric investigations are conducted to evaluate the effects of decision variables on the performances of the GT-MHR and combined cycles. The performances of these cycles are then optimized from the viewpoints of first law, second law and exergoeconomics. It is found that, combining the GT-MHR with ammonia–water cycle not only enhances the first and second law efficiencies of the GT-MHR, but also it improves the cycle performance from the exergoeconomic perspective. The results show that, when the optimization is based on the exergoeconomics, the unit cost of products is reduced by 5.4% in combining the two mentioned cycles. This is achieved with a just about 1% increase in total investment cost rate since the helium mass flow in the combined cycle is lower than that in the GT-MHR alone. - Highlights: • Application of exergetic cost theory to the combined GT-MHR/ammonia–water cycle. • Enhanced exergoeconomic performance for the combined cycle compared to the GT-MHR. • Comparable investment costs for the combined cycle and the GT-MHR alone

  20. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)

    2005-11-15

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: {sup 235}U, which represents the 20% of the fresh uranium, {sup 233}U, which is produced by the transmutation of fertile {sup 232}Th, and {sup 239}Pu, which is produced by the transmutation of fertile {sup 238}U. In order to compensate the depletion of {sup 235}U with the breeding of {sup 233}U and {sup 239}Pu, the quantity of fertile nuclides must be much larger than that one of {sup 235}U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of {sup 235}U. At the same time, the amount of {sup 235}U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k {sub eff} and mass

  1. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: 235 U, which represents the 20% of the fresh uranium, 233 U, which is produced by the transmutation of fertile 232 Th, and 239 Pu, which is produced by the transmutation of fertile 238 U. In order to compensate the depletion of 235 U with the breeding of 233 U and 239 Pu, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of 235 U. At the same time, the amount of 235 U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k eff and mass evolution, reaction rates, neutron flux and spectrum at the

  2. Studies on the feasibility of the LWRs waste-thorium in-core fuel cycle in the Gas Turbine-Modular Helium Reactor

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233 U by fertile thorium. Since a mixture of pure 239 Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29eV of the fission and capture microscopic cross sections of 239 Pu diminishes also the absorption rate of fertile 232 Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232 Th is about 80-100 times smaller than the fission one of 239 Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233 U represents only a small fraction of the depleted 239 Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one. (author)

  3. Development history of the gas turbine modular high temperature reactor

    International Nuclear Information System (INIS)

    Brey, H.L.

    2001-01-01

    The development of the high temperature gas cooled reactor (HTGR) as an environmentally agreeable and efficient power source to support the generation of electricity and achieve a broad range of high temperature industrial applications has been an evolutionary process spanning over four decades. This process has included ongoing major development in both the HTGR as a nuclear energy source and associated power conversion systems from the steam cycle to the gas turbine. This paper follows the development process progressively through individual plant designs from early research of the 1950s to the present focus on the gas turbine modular HTGR. (author)

  4. Refined finite element modelling for the vibration analysis of large rotating machines: Application to the gas turbine modular helium reactor power conversion unit

    International Nuclear Information System (INIS)

    Combescure, D.; Lazarus, A.; Lazarus, A.

    2008-01-01

    This paper is aimed at presenting refined finite element modelling used for dynamic analysis of large rotating machines. The first part shows an equivalence between several levels of modelling: firstly, models made of beam elements and rigid disc with gyroscopic coupling representing the position of the rotating shaft in an inertial frame; secondly full three-dimensional (3D) or 3D shell models of the rotor and the blades represented in the rotating frame and finally two-dimensional (2D) Fourier model for both rotor and stator. Simple cases are studied to better understand the results given by analysis performed using a rotating frame and the equivalence with the standard calculations with beam elements. Complete analysis of rotating machines can be performed with models in the frames best adapted for each part of the structure. The effects of several defects are analysed and compared with this approach. In the last part of the paper, the modelling approach is applied to the analysis of the large rotating shaft part of the power conversion unit of the GT-MHR nuclear reactor. (authors)

  5. Refined finite element modelling for the vibration analysis of large rotating machines: Application to the gas turbine modular helium reactor power conversion unit

    Energy Technology Data Exchange (ETDEWEB)

    Combescure, D.; Lazarus, A. [CEA Saclay, DEN/DM2S/SEMT/DYN, Dynam Anal Lab, Saclay, (France); Lazarus, A. [Ecole Polytech, Mecan Solides Lab, F-91128 Palaiseau, (France)

    2008-07-01

    This paper is aimed at presenting refined finite element modelling used for dynamic analysis of large rotating machines. The first part shows an equivalence between several levels of modelling: firstly, models made of beam elements and rigid disc with gyroscopic coupling representing the position of the rotating shaft in an inertial frame; secondly full three-dimensional (3D) or 3D shell models of the rotor and the blades represented in the rotating frame and finally two-dimensional (2D) Fourier model for both rotor and stator. Simple cases are studied to better understand the results given by analysis performed using a rotating frame and the equivalence with the standard calculations with beam elements. Complete analysis of rotating machines can be performed with models in the frames best adapted for each part of the structure. The effects of several defects are analysed and compared with this approach. In the last part of the paper, the modelling approach is applied to the analysis of the large rotating shaft part of the power conversion unit of the GT-MHR nuclear reactor. (authors)

  6. Comparative studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 data libraries on the Monte Carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232 Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239 Pu, 233 U and 235 U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B. (author)

  7. The Gas Turbine Modular Helium Reactor Optimum Design for Instrumentation and Control

    International Nuclear Information System (INIS)

    Quinn, Edward 'Ted'; Rodriguez, Carmelo

    2002-01-01

    Nuclear power plants rely on instrumentation and control (I and C) systems for monitoring, control and protection. While the earlier generations of nuclear plants were built with largely analog instrumentation systems, the industrial base has moved to digital based system. The reason for this transition to digital I and C systems lies in their important advantage over existing analog systems. In essence, the challenge is to prove that the new digital guidelines included in NUREG 0800, Standard Review Plan Chapter 7 and all associated documents provided adequate direction and safety review for an advanced control design for a new nuclear plant in the United States. This paper focuses on how one of the new designs incorporates the latest technology and meets the new NRC requirements. (authors)

  8. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  9. Hydrogen Process Coupling to Modular Helium Reactors

    International Nuclear Information System (INIS)

    Shenoy, Arkal; Richards, Matt; Buckingham, Robert

    2009-01-01

    The U.S. Department of Energy (DOE) has selected the helium-cooled High Temperature Gas-Cooled Reactor (HTGR) as the concept to be used for the Next Generation Nuclear Plant (NGNP), because it is the most advanced Generation IV concept with the capability to provide process heat at sufficiently high temperatures for production of hydrogen with high thermal efficiency. Concurrently with the NGNP program, the Nuclear Hydrogen Initiative (NHI) was established to develop hydrogen production technologies that are compatible with advanced nuclear systems and do not produce greenhouse gases. The current DOE schedule for the NGNP Project calls for startup of the NGNP plant by 2021. The General Atomics (GA) NGNP pre-conceptual design is based on the GA Gas Turbine Modular Helium Reactor (GT-MHR), which utilizes a direct Brayton cycle Power Conversion System (PCS) to produce electricity with a thermal efficiency of 48%. The nuclear heat source for the NGNP consists of a single 600-MW(t) MHR module with two primary coolant loops for transport of the high-temperature helium exiting the reactor core to a direct cycle PCS for electricity generation and to an Intermediate Heat Exchanger (IHX) for hydrogen production. The GA NGNP concept is designed to demonstrate hydrogen production using both the thermochemical sulfur-iodine (SI) process and high-temperature electrolysis (HTE). The two primary coolant loops can be operated independently or in parallel. The reactor design is essentially the same as that for the GT-MHR, but includes the additional primary coolant loop to transport heat to the IHX and other modifications to allow operation with a reactor outlet helium temperature of 950 .deg. C (vs. 850 .deg. C for the GT-MHR). The IHX transfers a nominal 65 MW(t) to the secondary heat transport loop that provides the high-temperature heat required by the SI-based and HTE-based hydrogen production facilities. Two commercial nuclear hydrogen plant variations were evaluated with

  10. Helium production in reactor materials

    International Nuclear Information System (INIS)

    Lippincott, E.P.; McElroy, W.N.; Farrar, H. IV.

    1975-02-01

    Comparisons of integral helium production measurements with predictions based on ENDF/B Version IV cross sections have been made. It is concluded that an ENDF/B helium production cross section file should be established in order to ensure a complete and consistent cross section evaluation to meet accuracies required for LMFBR, CTR, and LWR applications. (U.S.)

  11. Deep-Burn Modular Helium Reactor Fuel Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  12. The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

    2004-01-01

    We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of {sup 239}Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

  13. The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw; Venneri, Francesco

    2004-01-01

    We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of 239 Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

  14. Helium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Longton, P.B.; Cowen, H.C.

    1975-01-01

    In helium cooled HTR's there is a by-pass circuit for cleaning purposes in addition to the main cooling circuit. This is to remove such impurities as hydrogen, methane, carbon monoxide and water from the coolant. In this system, part of the coolant successively flows first through an oxidation bed of copper oxide and an absorption bed of silica gel, then through activated charcoal or a molecular sieve. The hydrogen and carbon monoxide impurities are absorbed and the dry gas is returned to the main cooling circuit. To lower the hydrogen/water ratio without increasing the hydrogen fraction in the main cooling circuit, some of the hydrogen fraction converted into water is added to the cooling circuit. This is done, inter alia, by bypassing the water produced in the oxidation bed before it enters the absorption bed. The rest of the by-pass circuit, however, also includes an absorption bed with a molecular sieve. This absorbs the oxidized carbon monoxide fraction. In this way, such side effects as the formation of additional methane, carburization of the materials of the by-pass circuit or loss of graphite are avoided. (DG/RF) [de

  15. The problem of helium in structural materials for fusion reactor

    International Nuclear Information System (INIS)

    Nikiforov, A.S.; Zakharov, A.P.; Chuev, V.I.

    1982-01-01

    The processes of helium buildup in some metals and alloys at different energy neutron flux irradiation under thermonuclear reactor conditions are considered. The data on high temperature helium embrittlement of a number of stainless steels, titanium and aluminium alloys etc. are given A review of experiments concerning the implanted helium behaviour is presented. Possible reactions between helium atoms and point defects or their clusters are discussed. Analysed are material structure variations upon buildup in them up to 1 at % of helium

  16. Fuel and helium confinement in fusion reactors

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Attenberger, S.E.

    1993-01-01

    An expanded macroscopic model for particle confinement is used to investigate both fuel and helium confinement in reactor plasmas. The authors illustrate the relative effects of external sources of fuel, divertor pumping, and wall and divertory recycle on core, edge and scrape-off layer densities by using separate particle confinement times for open-quote core close-quote fueling (deep pellet or beam penetration, τ c ), open-quote shallow close-quote fueling (shallow pellet penetration or neutral atoms that penetrate the scrape-off layer, τ s ) and fueling in the scrape-off layer (τ sol ). Because τ s is determined by the parallel flow velocity and characteristic distance to the divertor plate, it can be orders of magnitude lower than either τ c or τ sol . A dense scrape-off region, desirable for reduced divertor erosion, leads to a high fraction of the recycled neutrals being ionized in the scrape-off region and poor core fueling efficiency. The overall fueling efficiency can then be dramatically improved with either shallow or deep auxillary fueling. Helium recycle is nearly always coupled to the scrape-off region and does not lead to strong core accumulation unless the helium pumping efficiency is much less than the fuel pumping efficiency, or the plasma preferentially retains helium over hydrogenic ions. Differences between the results of this model, single-τ p macroscopic models, and 1-D and 2-D models are discussed in terms of assumptions and boundary conditions

  17. The Modular Helium Reactor for Hydrogen Production

    International Nuclear Information System (INIS)

    E. Harvego; M. Richards; A. Shenoy; K. Schultz; L. Brown; M. Fukuie

    2006-01-01

    For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR

  18. Utilization of thorium in a Gas Turbine – Modular Helium Reactor

    International Nuclear Information System (INIS)

    Şahin, Hacı Mehmet; Erol, Özgür; Acır, Adem

    2012-01-01

    Highlights: ► Performance parameters for the original fuel in GT-MHR depending on time were found. ► A proper plutonium–thorium mixture ratio was found using the original fuel results. ► Performance comparison of plutonium mixture and original fuel was made. ► Comparison showed that weapons grade plutonium mixture can be used in the reactor. - Abstract: Gas Turbine-Modular Helium Reactor (GT-MHR) is one of the new types of the reactors with high efficiency and increased safety features. The usage of different kinds of fissile material in this reactor can increase the life of it. Weapons-grade plutonium (WGrPu), which can be acquired from the old dismantled nuclear weapons, can be an option in a GT-MHR. In order to increase the sustainability of the WGrPu resources this fuel can be mixed with thorium, which is a fertile material that can be found in the nature and has resources three times more than uranium. In this study, possibility of utilization of the weapons-grade plutonium–thorium mixture was investigated and an optimum mixture ratio was determined. The behavior of this mixture and the original fuel was studied by using MCNP5 1.4, Monteburns 2.0 and Origen 2.2 tools. Calculations showed that, a GT-MHR type reactor, which is using the original TRISO fuel particle mixture of 20% enriched uranium + natural uranium (original fuel) has an effective multiplication factor (k eff ) of 1.270. Corresponding to this k eff value the weapons grade plutonium/thorium oxide mixture was found 19%/81%. By using Monteburns Code, the operation time, which describes the time passed until the reactor reaches a k eff value of 1.02, was found as 515 days for the original fuel and 1175 days for the weapons grade plutonium mixture. Furthermore, the burn-up values for the original fuel and WGrPu fuels were found as 47.69 and 119.27 GWd/MTU, respectively.

  19. Hydrogen generation using the modular helium reactor

    International Nuclear Information System (INIS)

    Richards, M.; Shenoy, A.

    2004-01-01

    Process heat from a high-temperature nuclear reactor can be used to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850 deg.C to 950 deg.C can drive the sulfur-iodine (SI) thermochemical process to produce hydrogen with high efficiency. Electricity can also be used to split water, using conventional, low-temperature electrolysis. An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyser to generate hydrogen. In this paper we investigate the coupling of the Modular Helium Reactor (MHR) to the SI process and HTE. These concepts are referred to as the H2-MHR. Optimization of the MHR core design to produce higher coolant outlet temperatures is also discussed. The use of fixed orifices to control the flow distribution is a promising design solution for increasing the coolant outlet temperature without increasing peak fuel temperatures significantly

  20. Helium effect on mechanical property of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Yamamoto, Norikazu; Chuto, Toshinori; Murase, Yoshiharu; Nakagawa, Johsei

    2004-01-01

    High-energy neutrons produced in fusion reactor core caused helium in the structural materials of fusion reactors, such as blankets. We injected alpha particles accelerated by the cyclotron to the samples of martensite steel (9Cr3WVTaB). Equivalent helium doses injected to the sample is estimated to be up to 300 ppm, which were estimated to be equivalent to helium accumulation after the 1-year reactor operation. Creep tests of the samples were made to investigate helium embrittlement. There were no appreciable changes in the relation between the stresses and the rupture time, the minimum creep rate and the applied stress. Grain boundary effect by helium was not observed in ruptured surfaces. Fatigue tests were made for SUS304 samples, which contain helium up to 150 ppm. After 0.05 Hz cyclic stress tests, it was shown that the fatigue lifetime (cycles to rupture and extension to failure) are 1/5 in 150 ppm helium samples compared with no helium samples. The experimental results suggest martensite steel is promising for structural materials of fusion reactors. (Y. Tanaka)

  1. Reactor helium system, design specification, operation and handling

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-06-01

    Apart from detailed design specification of the helium cover gas system of the Ra reactor, this document includes description of the operating regime, instructions for manipulations in the system with the aim of achieving and maintaining stationary gas circulation [sr

  2. Helium production in mixed spectrum reactor-irradiated pure elements

    International Nuclear Information System (INIS)

    Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

    1986-01-01

    The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

  3. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  4. PG-100 helium loop in the MR reactor

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Yakovlev, V.V.; Tikhonov, N.I.

    1983-01-01

    Main systems and production equipment units of PG-100 helium loop in the MR reactor are described. Possible long-term synchronizing operation of loop and reactor as well as possibility of carrying out life-time tests of spherical fuel elements and materials are shown. Serviceability of spherical fuel elements under conditions similar to the ones of HTGR-50 operation as well as high serviceability of cleanup system accepted for HTGR are verified. Due to low radiation dose the loop is operated without limits, helium losses in the loop don't exceed 0.5%/24 h, taking account of experimental gas sampling

  5. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  6. Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Chikara; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

  7. Modular helium reactor for non-electric applications

    International Nuclear Information System (INIS)

    Shenoy, A.

    1997-01-01

    The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

  8. Coupled CFD - system-code simulation of a gas cooled reactor

    International Nuclear Information System (INIS)

    Yan, Yizhou; Rizwan-uddin

    2011-01-01

    A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

  9. Weld repair of helium degraded reactor vessel material

    International Nuclear Information System (INIS)

    Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A.; Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G.

    1990-01-01

    Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces

  10. Helium turbomachine design for GT-MHR power plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Orlando, R.J.

    1994-07-01

    The power conversion system in the gas turbine modular helium reactor (GT-MHR) power plant is based on a highly recuperated closed Brayton cycle. The major component in the direct cycle system is a helium closed-cycle gas turbine rated at 286 MW(e). The rotating group consists of an intercooled helium turbocompressor coupled to a synchronous generator. The vertical rotating assembly is installed in a steel vessel, together with the other major components (i.e., recuperator, precooler, intercooler, and connecting ducts and support structures). The rotor is supported on an active magnetic bearing system. The turbine operates directly on the reactor helium coolant, and with a temperature of 850 degree C (1562 degree F) the plant efficiency is over 47%. This paper addresses the design and development planning of the helium turbomachine, and emphasizes that with the utilization of proven technology, this second generation nuclear power plant could be in service in the first decade of the 21st century

  11. Study on fundamental features of helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2004-01-01

    The High temperature gas-cooled reactor (HTGR) coupled with helium turbine cycle is considered as one of the leading candidates for future nuclear power plants. The HTGR helium turbine cycle was analyzed and optimized. Then the focal point of investigation was concentrated on the fundamental thermodynamic and aerodynamic features of helium turbomachine. As a result, a helium turbomachine is different from a general combustion gas turbine in two main design features, that is a helium turbomachine has more blade stages and shorter blade length, which are caused by the helium property and the high pressure of a closed cycle, respectively. (authors)

  12. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

    Lockett, G.E.; Kinkead, A.N.

    1967-02-01

    In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

  13. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  14. Compatibility of vanadium alloys with reactor-grade helium for fusion reactor applications

    International Nuclear Information System (INIS)

    Bell, G.E.C.; Bishop, P.S.

    1993-01-01

    Tests were conducted to determine the compatibility of vanadium alloys with reactor-grade helium and to define the helium gas chemistry requirements for fusion reactors, miniature tensile specimens of V-5Cr-5Ti. V-10Cr-5Ti, and V-12.5Cr-5 Ti were exposed in a once-through system to helium with 70 vppm-H 2 (measured oxygen partial pressures of 10 -12 atm) and bottle helium (measured oxygen partial pressures of -4 atm) between 500 and 700 degree C for up to 1008 h. The weight changes in the specimens were recorded. The helium-exposed specimens were tensile tested, and the effects of exposure on mechanical properties were assessed. Exposure between 500 and 700 degree C for 1008 h in He+70 vppm-H 2 resulted in complete embrittlement of all the alloys in room temperature tensile tests. The fracture mode was primarily cleavage, probably caused by a hydrogen-induced shift in the ductile to brittle transition temperature (DBTT). Weight gains increased with temperature and were largest for the V-5Cr-5Ti alloy. Specimens exposed for 531 h between 500 and 700 degree C in bottle He exhibited two distinct fracture morphologies on the fracture surfaces. Brittle cleavage around the edges of specimens gave way to ductile dimpling in the center of the specimens. The brittle region around the periphery of the specimen is most likely the highest vanadium oxide. V 2 O 5

  15. High temperature helium-cooled fast reactor (HTHFR)

    International Nuclear Information System (INIS)

    Karam, R.A.; Blaylock, Dwayne; Burgett, Eric; Mostafa Ghiaasiaan, S.; Hertel, Nolan

    2006-01-01

    Scoping calculations have been performed for a very high temperature (1000 o C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam-UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing

  16. USE OF THE MODULAR HELIUM REACTOR FOR HYDROGEN PRODUCTION

    International Nuclear Information System (INIS)

    SCHULTZ, K.R.

    2003-01-01

    OAK-B135 A significant ''Hydrogen Economy'' is predicted that will reduce our dependence on petroleum imports and reduce pollution and greenhouse gas emissions. Hydrogen is an environmentally attractive fuel that has the potential to displace fossil fuels, but contemporary hydrogen production is primarily based on fossil fuels. The author has recently completed a three-year project for the US Department of Energy (DOE) whose objective was to ''define an economically feasible concept for production of hydrogen, using an advanced high-temperature nuclear reactor as the energy source''. Thermochemical water-slitting, a chemical process that accomplishes the decomposition of water into hydrogen and oxygen, met this objective. The goal of the first phase of this study was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen, and to select one for further detailed consideration. They selected the Sulfur-Iodine cycle. In the second phase, they reviewed all the basic reactor types for suitability to provide the high temperature heat needed by the selected thermochemical water splitting cycle and chose the helium gas-cooled reactor. In the third phase they designed the chemical flowsheet for the thermochemical process and estimated the efficiency and cost of the process and the projected cost of producing hydrogen. These results are summarized in this report

  17. An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

    International Nuclear Information System (INIS)

    Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

    2006-01-01

    The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

  18. Coupling the modular helium reactor to hydrogen production processes

    International Nuclear Information System (INIS)

    Richards, M.B.; Shenoy, A.S.; Schultz, K.R.

    2004-01-01

    Steam reforming of natural gas (methane) currently produces the bulk of hydrogen gas used in the world today. Because this process depletes natural gas resources and generates the greenhouse gas carbon dioxide as a by-product, there is a growing interest in using process heat and/or electricity generated by nuclear reactors to generate hydrogen by splitting water. Process heat from a high temperature nuclear reactor can be used directly to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850 deg C to 950 deg C can drive the sulphur-iodine (S-I) thermochemical process to produce hydrogen with high efficiency. The S-I process produces highly pure hydrogen and oxygen, with formation, decomposition, regeneration, and recycle of the intermediate chemical reagents. Electricity can also 1)e used directly to split water, using conventional, low-temperature electrolysis (LTE). Hydrogen can also be produced with hybrid processes that use both process heat and electricity to generate hydrogen. An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyzer to generate hydrogen. This process is of interest because the efficiency of electrolysis increases with temperature. Because of its high temperature capability, advanced stage of development relative to other high-temperature reactor concepts, and passive-safety features, the modular helium reactor (MHR) is well suited for producing hydrogen using nuclear energy. In this paper we investigate the coupling of the MHR to the S-I process, LTE, and HTE. These concepts are referred to as the H2-MHR. (author)

  19. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  20. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  1. Helium generation in fusion-reactor materials. Progress report, October-December 1982

    International Nuclear Information System (INIS)

    Kneff, D.W.; Farrar, H. IV.

    1982-01-01

    The objectives of this work are to measure helium generation rates of materials for Magnetic Fusion Reactor applications in the Be(d,n) neutron environment, to characterize this neutron environment, and to develop helium accumulation neutron dosimeters for routine neutron fluence and energy spectrum measurements in Be(d,n) and Li(d,n) neutron fields

  2. Thermal insulation of the high-temperature helium-cooled reactors

    International Nuclear Information System (INIS)

    Kharlamov, A.G.; Grebennik, V.N.

    1979-01-01

    Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper

  3. A high temperature reactor could be used to eliminate the Russian military plutonium

    International Nuclear Information System (INIS)

    Foucher, N.

    1999-01-01

    The GT-MHR reactor (Gas Turbine Modular Helium Reactor) aims the double objective to eliminate the Russian plutonium coming from weapons, ( until 3 tons by year) and to produce a competitive energy from a small-scale power reactor with a nuclear fuel that can be of different type (plutonium or uranium). This reactor has several advantages: a high yield (47%) as every high temperature reactor and to be used in combined cycle, a high level of safety because of its ability to evacuate the residual power in a totally passive way and because of the nature of its fuel that is made of ceramics with a very high melting point that is to say no possibility of core melt. The fission products are contained in the ceramics so that reactor cannot disseminate radioactivity in its structure and consequently does not induce irradiation for the personnel. (N.C.)

  4. Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

    International Nuclear Information System (INIS)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1981-01-01

    Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700 0 C to fluence of 3 x 10 26 n/m 2 , E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700 0 C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations

  5. Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1981-01-01

    Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700/sup 0/C to fluence of 3 x 10/sup 26/ n/m/sup 2/, E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700/sup 0/C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations.

  6. Correlation for boron carbide helium release in fast reactors

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Pitner, A.L.

    1977-04-01

    An empirical helium correlation for the helium release from boron carbide has been developed. The correlation provides a good fit to the experimental data in the temperature range from 800 to 1350 0 K, and burnup levels up to 80 x 10 20 captures/cm 3 . The correlation has the capability of extrapolation to 2200 0 K (3500 0 F) and 200 x 10 20 captures/cm 3 . In this range the helium release rate will not exceed the generation rate

  7. Isotopic alloying to tailor helium production rates in mixed spectrum reactors

    International Nuclear Information System (INIS)

    Mansur, L.K.; Rowcliffe, A.F.; Grossbeck, M.L.; Stoller, R.E.

    1985-01-01

    The purposes of this work are to increase the understanding of mechanisms by which helium affects microstructure and properties, to aid in the development of materials for fusion reactors, and to obtain data from fission reactors in regimes of direct interest for fusion reactor applications. Isotopic alloying is examined as a means of manipulating the ratio of helium transmutations to atom displacements in mixed spectrum reactors. The application explored is based on artificially altering the relative abundances of the stable isotopes of nickel to systematically vary the fraction of 58 Ni in nickel bearing alloys. The method of calculating helium production rates is described. Results of example calculations for proposed experiments in the High Flux Isotope Reactor are discussed

  8. Optimization of TRU burnup in modular helium reactor

    International Nuclear Information System (INIS)

    Yonghee, Kim; Venneri, F.

    2007-01-01

    An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

  9. Thermal and flow design of helium-cooled reactors

    International Nuclear Information System (INIS)

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  10. Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

    International Nuclear Information System (INIS)

    Yamamoto, Norikazu; Murase, Yoshiharu; Nagakawa, Johsei; Shiba, Kiyoyuki

    2001-01-01

    Mechanical response against transmutational helium production, alternatively susceptibility to helium embrittlement, in a nuclear fusion reactor was examined on 8Cr2WVTa martensitic steel, a prominent structural candidate for advanced fusion systems. In order to simulate DEMO (demonstrative) reactor environments, helium was implanted into the material at 823 K with concentrations up to 1000 appmHe utilizing an α-beam from a cyclotron. Creep rupture properties were subsequently determined at the same temperature and were compared with those of the material without helium. It has been proved that helium caused no meaningful deterioration in terms of both the creep lifetime and rupture elongation. Furthermore, failure occurred completely in a transgranular and ductile manner even after high concentration helium introduction and there was no symptom of grain boundary decohesion which very often arises in helium bearing materials. These facts would mirror preferable resistance of this steel toward helium embrittlement. (author)

  11. Back pressure helium leak testing of fuel elements for Dhruva research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, N G; Ahmad, Anis; Kulkarni, P G; Purushotham, D S.C. [Bhabha Atomic Research Centre, Bombay (India). Atomic Fuels Div.

    1994-12-31

    Leak tightness specification on fuel elements for reactor use is always very stringent. The fuel element fabricated for Dhruva reactor is specified to be leak-tight up to 1 x 10{sup -8} std. cc/sec. The fuel element consists of natural metallic uranium rod around 12.5 mm diameter and 3 meter long in encased in aluminium tube and seal welded at both ends. Since helium gas is not filled inside the fuel element while doing seal welding, the only way to do helium leak testing of such fuel rods is by back-pressure technique. This paper describes the development of test facility for carrying out such test and discusses the experiences of carrying out helium leak testing by back-pressure technique on more than 700 numbers of fuel rods for Dhruva reactor. (author). 4 refs., 3 figs., 1 tab.

  12. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Baldwin, D.H.

    1984-08-01

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  13. Design and analysis of helium Brayton power cycles for HiPER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sánchez, Consuelo, E-mail: csanchez@ind.uned.es [Dpto. Ingeniería Energética UNED, Madrid (Spain); Juárez, Rafael; Sanz, Javier [Dpto. Ingeniería Energética UNED, Madrid (Spain); Instituto de Fusión Nuclear/UPM, Madrid (Spain); Perlado, Manuel [Instituto de Fusión Nuclear/UPM, Madrid (Spain)

    2013-10-15

    Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER.

  14. Design and analysis of helium Brayton power cycles for HiPER reactor

    International Nuclear Information System (INIS)

    Sánchez, Consuelo; Juárez, Rafael; Sanz, Javier; Perlado, Manuel

    2013-01-01

    Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER

  15. Neutronics study on hybrid reactor cooled by helium, water and molten salt

    International Nuclear Information System (INIS)

    Li Zaixin; Feng Kaiming; Zhang Guoshu; Zheng Guoyao; Zhao Fengchao

    2009-01-01

    There is no serious magnetohydrodynamics (MHD) problem when helium,water or molten salt of Flibe flows in high magnetic field. Thus helium, water and Flibe were proposed as candidate of coolant for fusion-fission hybrid reactor based on magnetic confinement. The effect on neutronics of hybrid reactor due to coolant was investigated. The analyses of neutron spectra and fuel breeding of blanket with different coolants were performed. Variations of tritium breeding ratio (TBR), blanket energy multiplication (M) and keff with operating time were also studied. MCNP code was used for neutron transport simulation. It is shown that spectra change greatly with different coolants. The blanket with helium exhibits very hard spectrum and good tritium breeding ability. And fission reactions are mainly from fast neutron. The blanket with water has soft spectrum and high energy multiplication factor. However, it needs to improve TBR. The blanket with Flibe has hard spectrum and less energy release. (authors)

  16. The real gas behaviour of helium as a cooling medium for high-temperature reactors

    International Nuclear Information System (INIS)

    Hewing, G.

    1977-01-01

    The article describes the influence of the real gas behaviour on the variables of state for the helium gas and the effects on the design of high-temperature reactor plants. After explaining the basic equations for describing variables and changes of state of the real gas, the real and ideal gas behaviour is analysed. Finally, the influence of the real gas behaviour on the design of high-temperature reactors in one- and two-cycle plants is investigated. (orig.) [de

  17. Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

    International Nuclear Information System (INIS)

    Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

    2002-01-01

    A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

  18. Tensile properties and microstructure of helium injected and reactor irradiated V-20 Ti

    International Nuclear Information System (INIS)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1980-01-01

    The objective of this work was to determine the effect of preinjected helium followed by neutron irradiation on the mechanical properties and microstructure of V-20% Ti. These results will be used for the evaluation of the potential use of V-20% Ti in fusion reactor service

  19. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  20. Cryosorption of helium on argon frost in Tokamak Fusion Test Reactor neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.; Garzotto, V.; Dudek, L.E.; Grisham, L.R.; Martin, G.D.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.; Kim, J.

    1990-01-01

    Helium pumping on argon frost has been investigated on Tokamak Fusion Test Reactor (TFTR) neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, >20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a twofold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 Torr l of helium, the beamline base pressure rose to 2x10 -5 Torr from an initial value of 10 -8 Torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 Torr l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity

  1. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  2. Helium generation in fusion reactor materials. Technical progress report, April--September 1977

    International Nuclear Information System (INIS)

    1978-01-01

    The near-term objectives of this program are to measure the spectrum-integrated helium generation rates and cross sections of a number of pure elements and alloys in several high-intensity neutron sources, and to develop and demonstrate neutron dosimetry procedures using some of these materials. To this end, four neutron irradiation experiments have now been run: one using accelerator-produced d-Be neutrons, two using the accelerator-produced d-T reaction, and one in the neutron field of a mixed-spectrum fission reactor. All of these irradiations have incorporated a large number of helium-generation materials

  3. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  4. Adsorption purification of helium coolant of high-temperature gas-cooled reactors of carbon dioxide

    International Nuclear Information System (INIS)

    Varezhkin, A.V.; Zel'venskij, Ya.D.; Metlik, I.V.; Khrulev, A.A.; Fedoseenkin, A.N.

    1986-01-01

    A series experiments on adsorption purification of helium of CO 2 using national adsorbent under the conditions characteristic of HTGR type reactors cleanup system is performed. The experimnts have been conducted under the dynamic mode with immobile adsorbent layer (CaA zeolite) at gas flow rates from 0,02 to 0,055 m/s in the pressure range from 0,8 to 5 MPa at the temperature of 273 and 293 K. It is shown that the adsorption grows with the decrease of gas rate, i.e. with increase of contact time with adsorbent. The helium pressure, growth noticeably whereas the temperature decrease from 293 to 273 K results in adsorption 2,6 times increase. The conclusion is drawn that it is advisable drying and purification of helium of CO 2 to perform separately using different zeolites: NaA - for water. CaA - for CO 2 . Estimations of purification unit parameters are realized

  5. Preliminary study on helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Chen Yihua; Wang Jie; Zhang Zuoyi

    2003-01-01

    In the high temperature gas-cooled reactor (HTGR), gas turbine cycle is a new concept in the field of nuclear power. It combines two technologies of HTGR and gas turbine cycle, which represent the state-of-the-art technologies of nuclear power and fossil fuel generation respectively. This approach is expected to improve safety and economy of nuclear power plant significantly. So it is a potential scheme with competitiveness. The heat-recuperated cycle is the main stream of gas turbine cycle. In this cycle, the work medium is helium, which is very different from the air, so that the design features of the helium turbomachine and combustion gas turbomachine are different. The paper shows the basic design consideration for the heat-recuperated cycle as well as helium turbomachine and highlights its main design features compared with combustion gas turbomachine

  6. Considerations on the design of a helium circulator for a high temperature modular reactor system

    International Nuclear Information System (INIS)

    Dumm, K.; Donaldson, J.

    1988-01-01

    A modular helium cooled, high temperature reactor system with a thermal output of 200 MW per reactor has been developed by the KWU group for cogeneration of electricity and process steam. The flow of the reactor coolant - Helium at 60 bars and 250/700 deg. C is maintained by one circulator per reactor. The circulator is driven by a variable speed Siemens asynchronous motor and is submerged in the helium primary system. For operational reasons high reliability and availability of the circulator is required. The operational requirements for the circulator design are presented in this paper. The actual design has been carried out in close cooperation with the designer and manufacturer of all submerged circulators operating in AGR plants in Great Britain, James Howden Co. Renfrew, Scotland. Design solutions received so far and mainly based on sufficiently proven components - such as oil bath lubricated bearing systems - will be described. Special attention will be paid on the necessary test work; especially for the prototype to confirm the lay out. (author). 9 figs

  7. Evolution of criteria for repair work on helium lines of Cirus reactor

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Soni, R.S.; Kushwaha, H.S.

    2006-05-01

    The research reactor CIRUS uses light water as coolant and heavy water as moderator and is rated for a thermal power of 40 MW. This reactor has been in operation since 1960 and has undergone refurbishment work recently. In the CIRUS reactor, helium gas is utilised as the cover gas. The helium lines are connected with the tube sheet at the top of the calandria. There are eight such helium lines at the top of the calandria, out of which four are connected to one ring header, three to another ring header and the remaining one is single line. These helium gas lines have tongue and groove joints for connecting the stainless steel piping with the aluminium piping. With the prolonged operation of the plant, leakage was observed at these joints. As a part of reactor refurbishing work, these joints were required to be repaired. Since these joints are situated in an inaccessible area, the entire job was to be carried out remotely and therefore, a fail-safe scheme was to be evolved based on computer simulation and analytical work. The entire analysis work had many challenging aspects hence, utmost care was exercised while analytically formulating the scheme for the tightening of these flange joints by postulating the various possible scenarios and by maintaining the stress level within the limits, particularly at the fillet welds between the aluminium pipe and calandria tube sheet. Another challenging aspect of this job was to take care of various uncertainties regarding the prevailing status of the joints. This report highlights the methodology adopted to arrive at the optimum amount of tightening and sequence of tightening. This report also highlights how analytical simulation of actual site scenario was carried out based on site feedbacks at various stages of tightening operations and how strategies were formulated to overcome various challenges and also to take care of various uncertainties in the input information being reported by the site. The tightening work

  8. Accident tolerant high-pressure helium injection system concept for light water reactors

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

    Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

  9. First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

    International Nuclear Information System (INIS)

    Piping Supriatna

    2009-01-01

    The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

  10. Control of helium activity in the fuel reactor channels; Kontrola aktivnosti heliuma u tehnoloskim kanalima

    Energy Technology Data Exchange (ETDEWEB)

    Vidmar, M; Milosevic, M; Hadzic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1961-02-15

    The objective of this task was to study the possibility of detecting a damaged fuel channel, and to introduce automated procedure for continuous control of reactor channels during operation. The existing control systems at the RA reactor (permanent control of heavy water and helium activity, radiation monitoring of heavy water and helium system, measurements of fire damp gas percent) are not sufficient for fast detection of fuel element failures. Since a 'hot' fuel channel cannot be removed from the core because it should be cooled in the core by heavy water circulation, it is not possible to prevent contamination of heavy water by fission products. It is concluded that it is not indispensable to detect the failed fuel element promptly, i.e. that tome is not a critical issue.

  11. Method of collecting helium cover gas for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyamoto, Keiji; Ueda, Hiroshi.

    1981-01-01

    Purpose: To reduce the systematic facility cost in a heavy water moderated reactor by contriving the simplification of a helium cover gas collecting intake system. Method: A detachable low pressure metal tank and a neoprene balloon are prepared for a vacuum pump in a permanent vacuum drying facility. When all of the helium cover gas is collected from a heavy water moderated reactor, a large capacity of neoprene balloon capable of temporarily storing it under low pressure is connected to the exhaust of the vacuum pump. On the other hand, while the reactor is operating, a suitable amount of the low pressure tank or neoprene balloon is connected to the exhaust side of the pump, thereby regulating the pressure of the helium cover gas. When refeeding the cover gas, the balloon, with a large capacity for collecting and storing the cover gas is connected to the intake side of the pump. Thus, the pressure regulation, collection of all of the cover gas and refeeding of the cover gas can be conducted without using a high discharge pump and high pressure tank. (Kamimura, M.)

  12. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  13. Helium desorption in EFDA iron materials for use in nuclear fusion reactors

    International Nuclear Information System (INIS)

    Salazar R, A. R.; Pinedo V, J. L.; Sanchez, F. J.; Ibarra, A.; Vila, R.

    2015-09-01

    In this paper the implantation with monoenergetic ions (He + ) was realized with an energy of 5 KeV in iron samples (99.9999 %) EFDA (European Fusion Development Agreement) using a collimated beam, after this a Thermal Desorption Spectrometry of Helium (THeDS) was made using a leak meter that detects amounts of helium of up to 10 - - 12 mbar l/s. Doses with which the implantation was carried out were 2 x 10 15 He + /cm 2 , 1 x 10 16 He + /cm 2 , 2 x 10 16 He + /cm 2 , 1 x 10 17 He + /cm 2 during times of 90 s, 450 s, 900 s and 4500 s, respectively. Also, using the SRIM program was calculated the depth at which the helium ions penetrate the sample of pure ion, finding that the maximum distance is 0.025μm in the sample. For this study, 11 samples of Fe EFDA were prepared to find defects that are caused after implantation of helium in order to provide valuable information to the manufacture of materials for future fusion reactors. However understand the effects of helium in the micro structural evolution and mechanical properties of structural materials are some of the most difficult questions to answer in materials research for nuclear fusion. When analyzing the spectra of THeDS was found that five different groups of desorption peaks existed, which are attributed to defects of He caused in the material, these defects are He n V (2≤n≤6), He n V m , He V for the groups I, II and IV respectively. These results are due to the comparison of the peaks presented in the desorption spectrum of He, with those of other authors who have made theoretical calculations. Is important to note that the thermal desorption spectrum of helium was different depending on the dose with which the implantation of He + was performed. (Author)

  14. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  15. A helium-cooled blanket design of the low aspect ratio reactor

    International Nuclear Information System (INIS)

    Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

    1998-03-01

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

  16. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  17. An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

  18. A prestressed concrete pressure vessel for helium high temperature reactor system

    International Nuclear Information System (INIS)

    Horner, R.M.W.; Hodzic, A.

    1976-01-01

    A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

  19. CFD Analysis for Hot Spot Fuel Temperature of Deep-Burn Modular Helium Reactor

    International Nuclear Information System (INIS)

    Tak, Nam Il; Jo, Chang Keun; Jun, Ji Su; Kim, Min Hwan; Venneri, Francesco

    2009-01-01

    As an alternative concept of a conventional transmutation using fast reactors, a deep-burn modular helium reactor (DB-MHR) concept has been proposed by General Atomics (GA). Kim and Venneri published an optimization study on the DB-MHR core in terms of nuclear design. The authors concluded that more concrete evaluations are necessary including thermo-fluid and safety analysis. The present paper describes the evaluation of the hot spot fuel temperature of the fuel assembly in the 600MWth DB-MHR core under full operating power conditions. Two types of fuel shuffling scheme (radial and axial hybrid shuffling and axial-only shuffling) are investigated. For accurate thermo-fluid analysis, the computational fluid dynamics (CFD) analysis has been performed on a 1/12 fuel assembly using the CFX code

  20. Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

    2006-01-01

    Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

  1. Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

    International Nuclear Information System (INIS)

    Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

    2013-01-01

    The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

  2. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  3. GT-MHR as economical highly efficient inherently safe modular gas cooled reactor for electric power generation

    International Nuclear Information System (INIS)

    Baydakov, A.G.; Kodochigov, N.G.; Kuzavkov, N.G.; Vorontsov, V.E.

    2001-01-01

    The Gas Turbine Modular Helium Reactor (GT-MHR) project is a joint effort of organisations in the Russian Federation co-ordinated by the Ministry on Atomic Energy (MINATOM), General Atomics in the United States, Framatome in France, and Fuji Electric in Japan. This paper describes the structure, objectives and history of the GT-MHR project, as well as the plant design, including plant layout, components and fuel. It summarises the results of evaluations of the design safety characteristics and economics, for both a plutonium fueled single module and a uranium fueled four module plant. The plans and anticipated structure for an international project to construct the plant are also summarised. (author)

  4. High temperature reactor and helium turbine for naval propeller (Study of feasibility and performances of the system)

    International Nuclear Information System (INIS)

    Brisbois, J.; Malherbe, J.; Rastoin, J.; Courau; Metayer.

    1976-01-01

    The nuclear reactor HTGR can get an outlet helium temperature greater than 800 deg C. That gives the means to use an helium turbine in a direct cycle. This type of reactor has been studied for a supply-ship (25,000t, 85,000Cv) and feasibility of such a system can be proved without employing any unknown materials. Because the weakness of helium activity, only the core can be shielded. All the propeller system is inside a containment which has to stand with a small over pressure after a core depressurisation. An efficiency of 35% is realized in a compact set up. This nuclear propeller get a very long core life 860FPD with constant worth- and very flexible working conditions. The HTGR direct cycle make a naval propeller very attractive [fr

  5. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  6. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  7. Void nucleation by the helium atoms during lifetime of reactor pressure vessel

    International Nuclear Information System (INIS)

    Rahman, F.A.

    1984-01-01

    Void formation and growth has a great influence on the reactor pressure vessel steels during its lifetime and during post-irradiation annealing to increase its life. The present investigation aimed at the fact that if one can prevent void nucleation, accordingly one would not wary about void formation and growth. From that concept a model for helium production by transmutation reaction and corresponding swelling under irradiation conditions for several number of steels have been developed. This was done for recommending a steel type that can oppose such a phenomena. In the same time the present investigation gives a procedure utilizing such phenomena for checking the validity of pressure vessel steel used in the NPP

  8. High frequency way of helium ash removal from stellarator-reactor

    International Nuclear Information System (INIS)

    Grekov, D.L.

    2005-01-01

    The paper deals with the problem of helium ash removal from stellarator-reactor. The lower hybrid heating of ash ions is proposed to solve this problem. The theory of ion stochastic heating, developed earlier by Karney, is generalized on the case of heating in stellarators. The features of the lower hybrid waves propagation and the ions motion in the stellarator confining field are taken into account. With proper choice of wave parameters (such as frequency, antenna position and initial spectrum of longitudinal refractive index) the slow mode of LH waves penetrates from the launching system to plasma core (and back) without conversion to kinetic plasma mode or to fast mode. With all these going on, the LH wave is absorbed by alpha particles only. The electron Landau damping is negligibly small, and there is no bulk ions stochastic heating. The motion of high energy (>100 keV) ions in the LHD heliotron with inwardly shifted magnetic axis, as an example of stellarator type device, is calculated numerically using the single particle simulation code which couples modified Karney's ion stochastic heating theory. The effect of collisions was taken into account through the Monte Carlo equivalent of the Lorentz collision operator. It is shown, that due to interaction with lower hybrid wave, initially well-confined alpha particles are expelled from the plasma during the time period less then collision time. At the same time, the low hybrid heating does not remove the ions with energy higher than 500 keV. Therefore, it is possible to use this method of RF heating for helium ash removal in stellarator-reactor. The required LH power is estimated to be of the order of 10 MW. (author)

  9. The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

    International Nuclear Information System (INIS)

    Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

    2014-01-01

    The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

  10. High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

    International Nuclear Information System (INIS)

    Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

    2012-01-01

    Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

  11. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    International Nuclear Information System (INIS)

    Vaiana, F.

    2009-11-01

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  12. A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

    International Nuclear Information System (INIS)

    Baltazar, O.

    1978-01-01

    A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

  13. Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

    International Nuclear Information System (INIS)

    Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

    2008-01-01

    The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

  14. Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

    2008-01-01

    Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

  15. Recycling, inventory and permeation of hydrogen isotopes and helium in the first wall of a thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Gervasini, G.; Reiter, F.

    1989-01-01

    The work was divided into three parts. The first part, which is theoretical, examines the behaviour of hydrogen in metals. After an introduction on the presence of hydrogen isotopes in fusion reactors, the main phenomena connected with hydrogen-metal interaction are summarised: solubility, diffusivity and trapping in material defects. The metal temperature is highlighted as the main parameter in the description of the phenomena. The second part of the work, also theoretical, concerns the interaction between helium and metals. We have tried as much as possible to show analogies and differences in the comparisons of the behaviour of hydrogen. The main types of damage caused by helium in metallic structures, which are the most important consequence of helium-metal interaction, were summarised. The characteristics of helium were treated in greater depth than those of hydrogen, because the latter are very well known. Also, there is a vast literature on the hydrogen-metal interaction. In the third and last part of the work a model was identified which allows the simulation of the evolution of a system formed from a metal in which hydrogen and helium isotopes have been introduced. A system of algebraic-differential equations was used to study the temporal evolution of the concentrations, the recycling, the inventory and the permeation of tritium and helium considering that these atoms diffuse in the metallic lattice and remain trapped in the vacancies created inside the metal by the bombardment of the neutrons from the fusion reactions. For the numerical simulation a series of data intended to represent the situation inside a thermonuclear reactor as precisely as possible were used for the numerical simulation. Analysis of the system was preceded by the analytical resolution of the steady state equations so that they could be compared with the simulation results

  16. Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO{sub 2} gas

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yeon-Gun [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear and Energy Engineering, Jeju National University, 102 Jejudaehakno, Jeju-si 690-756, Jeju (Korea, Republic of); Park, Il-Woong [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Lee, Dong Won [Nuclear Fusion Engineering Development Center, Korea Atomic Energy Research Institute, Daedeokdaero 989 beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Eung-Soo, E-mail: kes7741@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2015-11-15

    Helium (He) cooling requires large circulation power to remove high heat from plasma side and nuclear heating by high energy neutron in fusion reactors due to its low density. Based on the recent findings that the heat transfer capability of the light gas can be enhanced by mixing another heavier gas, this study adds CO{sub 2} to a reference helium coolant and evaluates the cooling performance of the binary mixture for various compositions. To assess the cooling performance, computational fluid dynamic (CFD) analyses on the KO HCML (Korea Helium Cooled Molten Lithium) TBM are conducted. As a result, it is revealed that the binary mixing of helium, which has favorable thermophysical properties but the density, with a heavier noble gas or an unreactive gas significantly reduces the required circulation power by an order of magnitude with meeting the thermal design requirements. This is attributed to the fact that the density can be highly increased with small amount of a heavier gas while other gas properties are kept relatively comparable. The optimal CO{sub 2} mole fraction is estimated to be 0.4 and the circulation power, in this case, can be reduced to 13% of that of pure helium. This implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of the proposed binary mixing.

  17. Program for aerodynamic performance tests of helium gas compressor model of the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Kunimoto, Kazuhiko; Yan, Xing; Itaka, Hidehiko; Mori, Eiji

    2003-01-01

    Research and development program for helium gas compressor aerodynamics was planned for the power conversion system of the Gas Turbine High Temperature Reactor (GTHTR300). The axial compressor with polytropic efficiency of 90% and surge margin more than 30% was designed with 3-dimensional aerodynamic design. Performance and surge margin of the helium gas compressor tends to be lower due to the higher boss ratio which makes the tip clearance wide relative to the blade height, as well as due to a larger number of stages. The compressor was designed on the basis of methods and data for the aerodynamic design of industrial open-cycle gas-turbine. To validate the design of the helium gas compressor of the GTHTR300, aerodynamic performance tests were planned, and a 1/3-scale, 4-stage compressor model was designed. In the tests, the performance data of the helium gas compressor model will be acquired by using helium gas as a working fluid. The maximum design pressure at the model inlet is 0.88 MPa, which allows the Reynolds number to be sufficiently high. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  18. Low-cycle fatigue of heat-resistant alloys in high-temperature gas-cooled reactor helium

    International Nuclear Information System (INIS)

    Tsuji, H.; Kondo, T.

    1984-01-01

    Strain controlled low-cycle fatigue tests were conducted on four nickel-base heat-resistant alloys at 900 0 C in simulated high-temperature gas-cooled reactor (HTGR) environments and high vacuums of about 10 -6 Pa. The observed behaviors of the materials were different and divided into two groups when tests were made in simulated HTGR helium, while all materials behaved similarly in vacuums. The materials that have relatively high ductility and compatibility with impure helium at test temperature showed considerable resistance to the fatigue damage in impure helium. On the other hand, the alloys qualified with their high creep strength were seen to suffer from the adverse effects of impure helium and the trend of intergranular cracking as well. The results were analyzed in terms of their susceptibility to the environmentenhanced fatigue damage by examining the ratios of the performance in impure helium to in vacuum. The materials that showed rather unsatisfactory resistance were considered to be characterized by their limited ductility partly due to their coarse grain structure and susceptibility to intergranular oxidation. Moderate carburization was commonly noted in all materials, particularly at the cracked portions, indicating that carbon intrusion had occurred during the crack growth stage

  19. Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

    International Nuclear Information System (INIS)

    Silva, Clarysson Alberto da

    2009-01-01

    This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

  20. Contribution to the optimization of the coupling of nuclear reactors to desalination processes

    International Nuclear Information System (INIS)

    Dardour, S.

    2007-04-01

    This work deals with modelling, simulation and optimization of the coupling between nuclear reactors (PWR, modular high temperature reactors) and desalination processes (multiple effect distillation, reverse osmosis). The reactors considered in this study are PWR (Pressurized Water Reactor) and GTMHR (Gas Turbine Modular Helium Reactor). The desalination processes retained are MED (Multi Effect Distillation) and SWRO (Sea Water Reverse Osmosis). A software tool: EXCELEES of thermodynamic modelling of coupled systems, based on the Engineering Algebraic Equation Solver has been developed. Models of energy conversion systems and of membrane desalination processes and distillation have been developed. Based on the first and second principles of thermodynamics, these models have allowed to determine the optimal running point of the coupled systems. The thermodynamic analysis has been completed by a first economic evaluation. Based on the use of the DEEP software of the IAEA, this evaluation has confirmed the interest to use these types of reactors for desalination. A modelling tool of thermal processes of desalination in dynamic condition has been developed too. This tool has been applied to the study of the dynamics of an existing plant and has given satisfying results. A first safety checking has been at last carried out. The transients able to jeopardize the integrated system have been identified. Several measures aiming at consolidate the safety have been proposed. (O.M.)

  1. Optimization of the deep-burner-modular helium reactor (DB-MHR) concept for actinide incineration

    International Nuclear Information System (INIS)

    Trakas, Ch.; Bruna, G.B.

    2005-01-01

    The paper summarizes studies performed on the General Atomics Deep-Burner Modular Helium Reactor (DB-MHR) concept-design carried out by Framatome ANP, Areva's joint subsidiary with Siemens. Feasibility and sensitivity studies as well as fuel-cycle studies with probabilistic methodology are presented. Emphasis is put on most attractive physical and computational aspects of the concept. Current investigations on design uncertainties, the future search for ways to improve the transmutation value in a double-stratum strategy, and the computational tools improvement are also presented. The Areva HTR, ANTARES, uses a similar prismatic core design. In that context, we revisited and optimized the Deep Burn concept. Typical values of transmutation ratio were established at 70%. Accounting for reactivity control needs estimated at 300 pcm, the cycle length can be estimated at 480 full-power days, giving an overall fuel irradiation time (6 fuel cycles) of about 10 calendar year, assuming as usual an average capacity factor of the plant of 80%. This study indicates a quite significant (96%) achievement for fissile material incineration. After a 500 year cooling-time, radiotoxicity (without fission products) is reduced by a factor 4

  2. Development of helium turbines associated with high temperature nuclear reactors for electric power plants

    International Nuclear Information System (INIS)

    Chaboseau, J.

    1975-01-01

    First, the system is defined and found perfectly adapted, in the industrial meaning of the term, to the utilization with high temperature nuclear reactors. Subsequently, a realistic determination of its original feature and that of the turbine is tempted: these features are few, particularly if the existence of numerous, already utilized, direct cycle devices is considered. Then, the influence of main characteristic parameters is evaluated, i.e. the choice of thermodynamic cycle, the utilization of helium, the principle of the layout and the rating of the power station. It appears to be sure that the engineering designs are of great importance for the turbine designer. For the latter, however, the technical aspects of the required developments will be of the conventional type. The character and the importance of these developments will mainly depend on how the system is spread in industry. It seems possible that the reasonable choice of the data of the first generating station allows to reduce the preliminary developments to a minimum amount while ensuring the subsequent evolution of the techniques. The inherent research of perfectness of the system justifies to anticipate long-term developments which are pursued in parallel with the commissioning of subsequent generations of power stations [fr

  3. Dynamic simulation for scram of high temperature gas-cooled reactor with indirect helium turbine cycle system

    International Nuclear Information System (INIS)

    Li Wenlong; Xie Heng

    2011-01-01

    A dynamic analysis code for this system was developed after the mathematical modeling and programming of important equipment of 10 MW High Temperature Gas Cooled Reactor Helium Turbine Power Generation (HTR-10GT), such as reactor core, heat exchanger and turbine-compressor system. A scram accident caused by a 0.1 $ reactivity injection at 5 second was simulated. The results show that the design emergency shutdown plan for this system is safe and reasonable and that the design of bypass valve has a large safety margin. (authors)

  4. Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

    International Nuclear Information System (INIS)

    Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

    2005-01-01

    The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

  5. Study of helium behaviour in body-centered cubic structures for new nuclear reactor generations: experimental approach in well characterized materials

    International Nuclear Information System (INIS)

    Gorondy-Novak, Sofia Maria

    2017-01-01

    The presence of helium produced during the operation of future fast reactors and fusion reactors in core structural materials induces a deterioration of their mechanical properties (hardening, swelling, embrittlement). In order to pursue the development of the metallic structural alloys, it is necessary to comprehend the He interaction with the metal lattice thus the point in common is the study of the metallic components with body-centered cubic structure (bcc) of future alloys, such as iron and/or vanadium. Ion implantation of ions "4He was employed with the aim of simulating the damaging effects associated with the helium accumulation, the point defects' creation (vacancies, self-interstitials) and the He cluster formation in future reactors. Helium evolution in pure iron and pure vanadium has been revealed from the point of view of the trapping sites' nature and well as the helium migration mechanisms and the nucleation/growth of bubbles. These phenomena were studied by coupling different complementary techniques. Despite of the fact that some mechanisms involved seem to be similar for both bcc metals, the comparison between the helium behavior in iron and vanadium shows certain differences. Microstructural defects, including grain boundaries and implanted helium concentration (dose) in both bcc metals will play significant roles on the helium behavior at high temperature. The acquired experimental data coupled with simulation methods contribute to the future development in terms of kinetic and thermodynamic data management of helium behavior in the metal components of the alloys of nuclear interest. (author) [fr

  6. Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1994-06-01

    An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.

  7. Adsorption removal of carbon dioxide from the helium coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Varezhin, A.V.; Fedoseenkov, A.N.; Khrulev, A.A.; Metlik, I.V.; Zel venskii, Y.D.

    1986-01-01

    This paper conducts experiments on the removal of CO 2 from helium by means of a Soviet-made adsorbent under the conditions characteristic of high-temperature gas-cooled reactor cleaning systems. The adsorption of CO 2 from helium was studied under dynamic conditions with a fixed layer of adsorbent in a flow-through apparatus with an adsorber 16 mm in diameter. The analysis of the helium was carried out by means of a TVT chromatograph. In order to compare the adsorption of CO 2 on CaA zeolite under dynamic conditions from the helium stream under pressure with the equilibrium adsorption on the basis of pure CO 2 , the authors determined the adsorption isotherm at 293 K by the volumetric method over a range of CO 2 equilibrium pressures from 260 to 11,970 Pa. Reducing the adsorption temperature to 273 K leads to a considerable reduction in the energy costs for regeneration, owing to the increase in adsorption and the decrease in the number of regeneration cycles; the amount of the heating gas used is reduced to less than half

  8. Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

    International Nuclear Information System (INIS)

    Ball, S.J.; Nypaver, D.J.

    1994-01-01

    An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ''operator'' involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor's fuel or metallic components

  9. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  10. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  11. Design and optimization of a multistage turbine for helium cooled reactor

    International Nuclear Information System (INIS)

    Braembussche, R.A. van den; Brouckaert, J.F.; Paniagua, G.; Briottet, L.

    2008-01-01

    This paper describes the aerodynamic design and explores the performance limits of a 600 MWt multistage helium turbine for a high temperature nuclear reactor closed cycle gas turbine. The design aims for maximum performance while limiting the number of stages for reasons of rotor dynamics and weight. A first part discusses the arguments that allow a preliminary selection of the overall dimensions by means of performance prediction correlations and simplified stress considerations. The rotational speed being fixed at 3000 rpm, the only degrees of freedom for the design are: the impeller diameter, number of stages and stage loading. The optimum load distribution of the different stages, the main flow parameters and the blade overall dimensions are defined by means of a 2D through-flow analysis method. The resulting absolute and relative flow angles and span-wise velocity variation are the input for a first detailed design by an inverse method. The latter defines the different 2D blade sections corresponding to prescribed optimum velocity distributions. The final 3D blade definition is made by means of a computer based 3D-DESIGN system developed at the von Karman Institute. This method combines a 3D Navier-Stokes (NS) solver, Database, Artificial Neural Network and Genetic Algorithm into a two level optimization technique for compressor and turbine stages. The use of 3D Navier-Stokes solvers allows full accounting of the secondary flow losses and optimization of the compound leaning of the stator vanes. The performance of the individual stages is used to define the multistage operating curves. The last part of the paper describes an evaluation of the cooling requirements of the first turbine rotor

  12. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  13. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    International Nuclear Information System (INIS)

    Snead, Lance Lewis; Besmann, Theodore M.; Collins, Emory D.; Bell, Gary L.

    2010-01-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  14. Detail analysis of tritium permeation in the metal liquid channels of the regenerating sheaths of a fusion reactor in presence of helium bubbles

    International Nuclear Information System (INIS)

    Banet, L.; Mas de les Valls, E.; Sedano, L. A.

    2012-01-01

    Inside the channels of liquid metal of the fusion reactor regenerative wrappers, the possible existence of nucleated helium bubbles is not remote. Helium is formed joined the tritium in the escaped neutrons of plasma with lithium. The accumulation of helium in the contact surfaces, between the structure and ML, lead a reduction of heat transfer, at the same time a reduction in the permeation of tritium. The coexistence of three phases in touch: metal liquid, helium and structural material, makes the transport of heat and tritium in a complex phenomenon. To enrich tritium transport studies conducted in the past, there is now a detail analysis of the helium bubble environment adhered to the channel ML wall of a regenerative wrap. For the study we used a CFD tool development on free code OpenFOAM.

  15. Cryosorption of helium on argon frost TFTR [Tokamak Fusion Test Reactor] neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.

    1989-11-01

    Helium pumping on argon frost has been investigated on TFTR neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼ 25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, > 20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a two-fold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 torr-l of helium, the beamline base pressure rose to 2x10 -5 torr from an initial value of 10 -8 torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 torr-l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity. 29 refs., 7 figs

  16. Influence of implanted helium on nickel resistance under simulation of plasma flux disruption in nuclear fusion reactor

    International Nuclear Information System (INIS)

    Kadin, B.A.; Pol'skij, V.I.; Yakushin, V.L.; Markin, A.V.; Tserevitinov, S.S.; Vasil'ev, V.I.

    1992-01-01

    Investigation results are presented of radiation erosion of constructive materials of the first wall of a thermonuclear reactor. The erosion is conditioned by successive repeated action of pulse processes, imitating plasma disruption, and helium ion fluxes at 40 keV and 2 x 10 21 -10 22 m -2 fluence. As imitating processes are used fluxes of deuterium high-temperature plasma. It is shown that preliminary action by high-temperature plasma leads to substantial suppression of radiation erosion, included by subsequent ion irradiation

  17. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  18. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.; Hoffman, M.A.; Johnson, G.L.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  19. Key physical parameters and temperature reactivity coefficients of the deep burn modular helium reactor fueled with LWRs waste

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Cetnar, Jerzy E-mail: jerzy@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

    2004-11-01

    We investigated some important neutronic features of the deep burn modular helium reactor (DB-MHR) using the MCNP/MCB codes. Our attention was focused on the neutron flux and its spectrum, capture to fission ratio of {sup 239}Pu and the temperature coefficient of fuel and moderator. The DB-MHR is a graphite-moderated helium-cooled reactor proposed by General Atomic to address the need for a fast and efficient incineration of plutonium for non-proliferation purposes as well as the management of light water reactors (LWRs) waste. In fact, recent studies have shown that the use of the DB-MHR coupled to ordinary LWRs would keep constant the world inventory of plutonium for a reactor fleet producing 400 TW{sub e}/y. In the present studies, the DB-MHR is loaded with Np-Pu driver fuel (DF) with an isotopic composition corresponding to LWRs spent fuel waste. DF uses fissile isotopes (e.g. {sup 239}Pu and {sup 241}Pu), previously generated in the LWRs, and maintains criticality conditions in the DB-MHR. After an irradiation of three years, the spent DF is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). TF mainly contains non-fissile actinides which undergo neutron capture and transmutation during the subsequent three-year irradiation in the DB-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. After extraction of the spent TF, irradiated for three years, over 94% of {sup 239}Pu and 53% of all actinides coming from LWRs waste will have been destroyed in the DB-MHR. In this paper we look at the operation conditions at equilibrium for the DB-MHR and evaluate fluxes and reactivity responses using state of the art 3-D Monte Carlo simulations.

  20. Contribution to the optimization of the coupling of nuclear reactors to desalination processes; Contribution a l'optimisation du couplage des reacteurs nucleaires aux procedes de dessalement

    Energy Technology Data Exchange (ETDEWEB)

    Dardour, S

    2007-04-15

    This work deals with modelling, simulation and optimization of the coupling between nuclear reactors (PWR, modular high temperature reactors) and desalination processes (multiple effect distillation, reverse osmosis). The reactors considered in this study are PWR (Pressurized Water Reactor) and GTMHR (Gas Turbine Modular Helium Reactor). The desalination processes retained are MED (Multi Effect Distillation) and SWRO (Sea Water Reverse Osmosis). A software tool: EXCELEES of thermodynamic modelling of coupled systems, based on the Engineering Algebraic Equation Solver has been developed. Models of energy conversion systems and of membrane desalination processes and distillation have been developed. Based on the first and second principles of thermodynamics, these models have allowed to determine the optimal running point of the coupled systems. The thermodynamic analysis has been completed by a first economic evaluation. Based on the use of the DEEP software of the IAEA, this evaluation has confirmed the interest to use these types of reactors for desalination. A modelling tool of thermal processes of desalination in dynamic condition has been developed too. This tool has been applied to the study of the dynamics of an existing plant and has given satisfying results. A first safety checking has been at last carried out. The transients able to jeopardize the integrated system have been identified. Several measures aiming at consolidate the safety have been proposed. (O.M.)

  1. International project GT-MHR - New generation of nuclear reactors

    International Nuclear Information System (INIS)

    Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

    2001-01-01

    Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

  2. Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

    International Nuclear Information System (INIS)

    Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

    2014-01-01

    Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

  3. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core

    International Nuclear Information System (INIS)

    Belo, Allan Cavalcante

    2016-01-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4 th generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range and

  4. Preliminary study on application of Pd composite membrane in helium purification system of high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Cai Jianhua; Yang Xiaoyong; Wang Jie; Yu Suyuan

    2008-01-01

    Helium purification system (HPS) is the main part of the helium auxiliary system of high-temperature gas-cooled reactors (HTGR), also in fusion reactors. Some exploratory work was carried out on the application of Pd composite membrane in the separation of He and H 2 . A typical single stripper permeator with recycle (SSP) system was designed, based on the design parameters of a small scale He purification test system CIGNE in CADARACHE, CEA, France, and finite element analysis method was used to solve the model. The total length of membrane module is fixed to 0.5 m. The results show that the concentration of H 2 is found to reduce from 1 000 μL/L in feed gas to 5 μL/L in the product He (the upper limitation of HPS in HTGR). And the molar ratio of product He to feed gas is 96.18% with the optimized ratio of sweep gas to retentive gas 0. 3970. It's an exponential distribution of H 2 concentration along the membrane module. The results were also compared with the other two popular designs, two stripper in series permeator (TSSP) and continuous membrane column (CMC). (authors)

  5. Investigation of Countercurrent Helium-Air Flows in Air-ingress Accidents for VHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Xiaodong; Christensen, Richard; Oh, Chang

    2013-10-03

    The primary objective of this research is to develop an extensive experimental database for the air- ingress phenomenon for the validation of computational fluid dynamics (CFD) analyses. This research is intended to be a separate-effects experimental study. However, the project team will perform a careful scaling analysis prior to designing a scaled-down test facility in order to closely tie this research with the real application. As a reference design in this study, the team will use the 600 MWth gas turbine modular helium reactor (GT-MHR) developed by General Atomic. In the test matrix of the experiments, researchers will vary the temperature and pressure of the helium— along with break size, location, shape, and orientation—to simulate deferent scenarios and to identify potential mitigation strategies. Under support of the Department of Energy, a high-temperature helium test facility has been designed and is currently being constructed at Ohio State University, primarily for high- temperature compact heat exchanger testing for the VHTR program. Once the facility is in operation (expected April 2009), this study will utilize high-temperature helium up to 900°C and 3 MPa for loss-of-coolant accident (LOCA) depressurization and air-ingress experiments. The project team will first conduct a scaling study and then design an air-ingress test facility. The major parameter to be measured in the experiments is oxygen (or nitrogen) concentration history at various locations following a LOCA scenario. The team will use two measurement techniques: 1) oxygen (or similar type) sensors employed in the flow field, which will introduce some undesirable intrusiveness, disturbing the flow, and 2) a planar laser-induced fluorescence (PLIF) imaging technique, which has no physical intrusiveness to the flow but requires a transparent window or test section that the laser beam can penetrate. The team will construct two test facilities, one for high-temperature helium tests with

  6. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems. Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    International Nuclear Information System (INIS)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E.; Shenoy, A.; Baccaglini, G.; Pletcher, R.H.; Wallace, J.M.; Vukoslavcevic, P.; Jackson, J.D.; Kunugi, T.; Satake, S.-I.

    2002-01-01

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications

  7. Evaluation of materials for heat exchanging components in advanced helium-cooled reactors

    International Nuclear Information System (INIS)

    Schubert, F.

    1984-01-01

    The qualification of metallic materials for advanced HTR applications is based on creep behaviour, fatigue properties, structural stability and corrosion resistance. A brief state of the art is provided for the materials for heat exchanging components. The experimental results are treated with respect to the importance for the design, the characteristic of time-depend materials behaviour are evaluated. Of specific interest are the possible effects of helium on the mechanical properties. Helium, which serves as primary coolant, contains traces of reactive impurities such as hydrogen, methane, carbon monoxide and water vapor. The evaluation of the HTR materials program serves as basis for structural design rules of components with operation temperatures above 800 deg C. The materials mechanical topics are discussed. Alloy improvement and the progress in development of new alloys are reviewed. (author)

  8. Reactor irradiation and helium-3 effects on mechanical properties of alpha-titanium alloys

    International Nuclear Information System (INIS)

    Tebus, V.N.; Alekseev, Eh.F.; Golikov, I.V.

    1990-01-01

    Dependence of α-titanium alloy mechanical properties on test temperature and neutron fluence is investigated. Irradiation is shown to result in material hardening and in their plasticity reduction, but residual plasticity remains rather high. Additional reduction of plasticity results in helium-3 introduced in materials under irradiation. Restoration of properties is observed at test temperature higher 500 deg C. Irradiation by fast neutrons up to high fluences (1.4·10 23 cm -2 ) results in essential alloy softening

  9. Sensisivity and Uncertainty analysis for the Tritium Breeding Ratio of a DEMO Fusion reactor with a Helium cooled pebble bed blanket

    OpenAIRE

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2016-01-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design c...

  10. Prospects for application of high-temperature helium reactor (HTHR) to provide for power needs in refineries and petrochemical plants

    International Nuclear Information System (INIS)

    Feigin, E.A.; Raud, E.A.; Romanova, E.G.; Panasenko, P.A.; Nikitin, V.N.

    1990-01-01

    An analysis performed shows that heat supply from High Temperature Helium Reactor (HTHR) located several miles from crude oil refining units, operating at 360-400 deg. C, can be arranged to use organic heat-carriers. At higher operating temperatures most acceptable are saline carriers. However the final choice of heat carries requires extra research including that on a large scale basis in order to improve the technology and equipment of the heat-supply system. The following problems can be solved by implementing HTHRs at the crude oil processing and petrochemical plants: improving pollution control, making more hydrocarbon fuel available for other uses, intensifying the operation of process units, and making them less fire hazardous, increasing the power efficiency of process plants

  11. Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments

    International Nuclear Information System (INIS)

    Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

    1984-01-01

    In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700 0 C, but the creep tests at 800 0 C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X

  12. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  13. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  14. Supercritical CO2 Brayton power cycles for DEMO fusion reactor based on Helium Cooled Lithium Lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Herranz, Luis Enrique; Fernández, Iván; Cantizano, Alexis; Moratilla, Beatriz Yolanda

    2015-01-01

    Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles (S-CO 2 ) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO 2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of HCLL fusion reactor. • Low temperature sources have been successfully integrated with high temperature ones. • Optimization of thermal sources integration improves 5% the electricity production. • Assessment of pumping power with sources and sink loops results on 20% of gross power. • Matching of key parameters has conducted to 1/3 of reduction in heat

  15. Criticality assessment for prismatic high temperature reactors by fuel stochastic Monte Carlo modeling

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    Modeling of prismatic high temperature reactors requires a high precision description due to the triple heterogeneity of the core and also to the random distribution of fuel particles inside the fuel pins. On the latter issue, even with the most advanced Monte Carlo techniques, some approximation often arises while assessing the criticality level: first, a regular lattice of TRISO particles inside the fuel pins and, second, the cutting of TRISO particles by the fuel boundaries. We utilized two of the most accurate Monte Codes: MONK and MCNP, which are both used for licensing nuclear power plants in United Kingdom and in the USA, respectively, to evaluate the influence of the two previous approximations on estimating the criticality level of the Gas Turbine Modular Helium Reactor. The two codes exactly shared the same geometry and nuclear data library, ENDF/B, and only modeled different lattices of TRISO particles inside the fuel pins. More precisely, we investigated the difference between a regular lattice that cuts TRISO particles and a random lattice that axially repeats a region containing over 3000 non-cut particles. We have found that both Monte Carlo codes provide similar excesses of reactivity, provided that they share the same approximations.

  16. Gas-cooled fast-breeder reactor. Helium Circulator Test Facility updated design cost estimate

    International Nuclear Information System (INIS)

    1979-04-01

    Costs which are included in the cost estimate are: Titles I, II, and III Architect-Engineering Services; Titles I, II, and III General Atomic Services; site clearing, grading, and excavation; bulk materials and labor of installation; mechanical and electrical equipment with installation; allowance for contractors' overhead, profit, and insurance; escalation on materials and labor; a contingency; and installation of GAC supplied equipment and materials. The total estimated cost of the facility in As Spent Dollars is $27,700,000. Also included is a cost comparison of the updated design and the previous conceptual design. There would be a considerable penalty for the direct-cooled system over the indirect-cooled system due to the excessive cost of the large diameter helium loop piping to an outdoor heat exchanger. The indirect cooled system which utilizes a helium/Dowtherm G heat exchanger and correspondingly smaller and lower pressure piping to its outdoor air cooler proved to be the more economical of the two systems

  17. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    International Nuclear Information System (INIS)

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  18. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    International Nuclear Information System (INIS)

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  19. Experimental study on cryogenic adsorption of methane by activated carbon for helium coolant purification of High-Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Chang, Hua; Wu, Zong-Xin; Jia, Hai-Jun

    2017-01-01

    Highlights: • The cryogenic CH 4 adsorption on activated carbon was studied for design of HTGR. • The breakthrough curves at different conditions were analyzed by the MTZ model. • The CH 4 adsorption isotherm was fitted well by the Toth model and the D-R model. • The work provides valuable reference data for helium coolant purification of HTGR. - Abstract: The cryogenic adsorption behavior of methane on activated carbon was investigated for helium coolant purification of high-temperature gas-cooled reactor by using dynamic column breakthrough method. With helium as carrier gas, experiments were performed at −196 °C and low methane partial pressure range of 0–120 Pa. The breakthrough curves at different superficial velocities and different feed concentrations were measured and analyzed by the mass-transfer zone model. The methane single-component adsorption isotherm was obtained and fitted well by the Toth model and the Dubinin-Radushkevich model. The adsorption heat of methane on activated carbon was estimated. The cryogenic adsorption process of methane on activated carbon has been verified to be effective for helium coolant purification of high-temperature gas-cooled reactor.

  20. Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

  1. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    International Nuclear Information System (INIS)

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  2. Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hitoki, Shigehisa; Sugihara, Masayoshi; Saito, Seiji; Fujisawa, Noboru

    1985-01-01

    A detailed calculation of the behavior of fuel and He particles in tokamak reactor with pumped-limiter is performed by one-dimensional tokamak transport code. Energy of neutral particles flowing back from limiter chamber is calculated by two-dimensional Monte Carlo neutral code. Feasibility of He ash exhaust and heat removal by the pumped-limiter are analyzed. Following features of the pumped-limiter are clarified: (1) Electron temperature decays rapidly in radial direction in scrape-off layer, while density profile is broader than that of temperature. (2) Helium accumulation in main plasma can be kept at desired level by rather short limiter and moderate pumping system. (3) Minimum amount of tritium pumped out little depends on limiter length. (4) Although high temperature plasma in scrape-off layer could be realized by large pumping and ideal pellet injection, it is not sufficiently high to reduce the erosion of the limiter surface and the leading edge. In conclusion, He ash exhaust may be possible by the pumped-limiter, while the heat load and erosion will be so high that the pumped-limiter may not be applicable unless the boundary plasma is cooled by radiation or by some other means. (author)

  3. Alpha-particle confinement and helium ash accumulation in stellarator reactors

    International Nuclear Information System (INIS)

    Ho, D.D.-M.; Kulsrud, R.M.

    1986-01-01

    The effect of local magnetic wells produced by the external helical windings of a stellarator reactor on α-particle confinement is investigated by using the Fokker-Planck equation. It is found that α-particles can deposit most of their energy into the background plasma before they scatter into the trapped region in velocity space and are lost. An estimate is made on the steady-state, low energy (thermal) α-particle concentration. The result shows that this concentration will be less than a few per cent of the background plasma density. (author)

  4. Comparative studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 data libraries by monte carlo modeling of high temperature reactors on plutonium based fuel cycles

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw; Cetnar, Jerzy

    2004-01-01

    We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library. (author)

  5. Severe fuel damage in steam and helium environments observed in in-reactor experiments

    International Nuclear Information System (INIS)

    Saito, S.; Shiozawa, S.

    1984-01-01

    The bahavior of severe fuel damages has been studied in gaseous environments simulating core uncovery accidents in the in-reactor experiments utilizing the NSRR. Two types of cladding relocation modes, azimuthal flow and melt-down, were revealed through the parametric experiments. The azimuthal flow was evident in an oxidizing environment in case of no oxide film break. The melt-down can be categorized into flow-down and move-down, according to the velocity of the melt-down. Cinematographies showed that the flow-down was very fast as water flows down while the move-down appeared to be much slower. The flow-down was possible in an unoxidizing environment, whereas the move-down of molten cladding occured through a crack induced in an oxide film in an oxidizing environment. The criterion of the relocation modes was developed as a function of peak cladding temperature and oxidation condition. It was also found that neither immediate quench nor fuel fracture occurred upon flooding when cladding temperature was about 1800 0 C at water injection. The external mechanical force is needed for fuel fracture. (orig.)

  6. Gas Turbine High Temperature Gas (Helium) Reactor Using Pebble Bed Fuel Derived from Spent Fuel

    International Nuclear Information System (INIS)

    Cole, Quentin

    2013-01-01

    Project goals: Build on the $1B investment spent during the NGNP Project for the only true Inherently Safe Small Modular Reactor Design – the only SMR design that can make this claim due to negative temperature coefficient of reactivity - no containment required – less construction cost. NPMC in Partnership with Pebble Bed Modular Group, a fully owned subsidiary of Eskom, RSA to Factory Build Complete Plant in Modular Sections at Factory Site in Oswego, NY for transport to site by rail or shipping for world wide export. NPMC will provide Project and Construction Management of all new builds from plant sites through construction, commissioning and startup using local labor. License and Construct ion of spent fuel processing facility in both NY and South Africa using Proven Technology. Ultimate goals of project: 1. Award of the 2013 US DOE Innovative SMR $452M cost share grant for US NRC License Certification 2.Build Full Scale Demonstration Plant at Koeburg, RSA with World Bank Funding managed by NPMC in collaboration with our legal firm, Haynes and Boone LLP 3. Take Plant Orders Immediately (10% Down Payment) 4. Form Strategic Alliance with Domestic and/or International Utility

  7. Potential applications of helium-cooled high-temperature reactors to process heat use

    International Nuclear Information System (INIS)

    Gambill, W.R.; Kasten, P.R.

    1981-01-01

    High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes

  8. Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

    International Nuclear Information System (INIS)

    Quadakkers, W.J.

    1985-01-01

    The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

  9. Purification by molecular sieve of helium used as inert cover gas in nuclear reactors; Epuration de l'helium de couverture des reacteurs nucleaires par adsorption sur tamis moleculaire

    Energy Technology Data Exchange (ETDEWEB)

    Rozenberg, J; Kahan, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    A method carried out at fairly low temperatures (between -50 and -80 deg. C) has been studied for the purification of the helium used as cover gas for heavy water in reactors. The use of the 5A molecular sieve has been adopted because of its superiority over other adsorbents in this temperature range. The particular problems connected with adsorption under dynamic conditions have been dealt with separately. The nitrogen adsorption isotherms have been plotted and the heat of adsorption calculated. (authors) [French] Une methode d'epuration, a temperature moderement basse (comprise entre -50 et -80 deg. C) de l'helium servant de couverture inerte a l'eau lourde des reacteurs a ete etudiee. L'emploi au tamis moleculaire 5A a ete retenu pour la superiorite de celui-ci sur d'autres adsorbants dans ce domaine de temperatures. Les problemes particuliers a l'adsorption en regime dynamique ont ete separement traites. Les isothermes d'adsorption d'azote ont ete tracees et la chaleur d'adsorp. tion calculee. (auteurs)

  10. Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

    International Nuclear Information System (INIS)

    Chow, J.G.Y.; Soo, P.; Epel, L.

    1978-01-01

    A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

  11. Research on dynamics and experiments about auxiliary bearings for the helium circulator of the 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Zhao, Yulan; Yang, Guojun; Liu, Xingnan; Shi, Zhengang; Zhao, Lei

    2016-01-01

    Highlights: • The research in this paper is based on the AMB helium circulator of HTR-10. • The dynamic rotor performance is analyzed by processing experimental data. • The mechanical bearing without lubrication can be applied in the HTR-10 system. - Abstract: The 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University. The auxiliary bearing is utilized in this system to meet particular requirements for the reactor. The main role of the auxiliary bearing is to constrain rotor displacements and also to support the rotor when the rotor drops down, which is caused by the active magnetic bearing (AMB) failure. The auxiliary bearing needs to endure huge impact, rapid angular acceleration and thermal shock. On the one hand, complex geometrical constructions and forces applied on the system bring difficulties and restrictions to establish an appropriate model to reveal the actual dynamic process. On the other hand, large volumes of data obtained from experiments show velocities and displacements of the rotor during the rotor drop process and then can indicate the actual dynamic interactions to a great extent. The research in this paper is based on the test rig of the AMB helium circulator of HTR-10. This paper aims to analyze the dynamic performance and contact forces of the rotor by processing experimental data. A measurement to estimate forces developed due to impacts of the rotor and the auxiliary bearings is presented. It is of great significance and provides certain foundation to elaborate the rotor drop process for the AMB helium circulator of HTR-10.

  12. Phenomenological studies and modelling of the gaseous impurities oxidation and adsorption mechanisms in helium: application for the purification system optimization in gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Legros, F.

    2008-01-01

    In GEN IV studies on future fission nuclear reactors, two concepts using helium as a coolant have been selected: GFR and VHTR. Among radioactive impurities and dusts, helium can contain H 2 , CO, CH 4 , CO 2 , H 2 O, O 2 , as well as nitrogenous species. To optimize the reactor functioning and lifespan, it is necessary to control the coolant chemical composition using a dedicated purification system. A pilot designed at the CEA allows studying this purification system. Its design includes three unit operations: H 2 and CO oxidation on CuO, then two adsorption steps. This study aims at providing a detailed analysis of the first and second purification steps, which have both been widely studied experimentally at laboratory scale. A first modelling based on a macroscopic approach was developed to represent the behaviour of the reactor and has shown that the CuO fixed bed conversion is dependent on the chemistry (mass transfer is not an issue) and is complete. The results of the structural analysis of the solids allow considering the CuO as particles made of 200 nm diameter grains. Hence, a new model at grain scale is proposed. It is highlighted that the kinetic constants from these two models are related with a scale factor which depends on geometry. A competition between carbon monoxide and hydrogen oxidation has been shown. Activation energies are around 30 kJ.mol-1. Simulation of the simultaneous oxidations leads to consider CO preferential adsorption. A similar methodology has been applied for CO 2 and H 2 O adsorption. The experimental isotherms showed a Langmuir type adsorption. Using this model, experimental and theoretical results agree. (author) [fr

  13. Helium behaviour in aluminium under hydrostatic pressure

    International Nuclear Information System (INIS)

    Sokurskij, Yu.N.; Tebus, V.N.; Zudilin, V.A.; Tumanova, G.M.

    1989-01-01

    Effect of hydrostatic compression on equilibrium helium bubbles in low aluminium-lithium alloy irradiated in reactor at 570 K is investigated. Measurements of hydrostatic density and electron-microscopic investigations have shown, that application of up to 2 GPa pressure reduces equilibrium size of helium bubbles and reduces helium swelling. Kinetics and thermodynamics of the process are considered with application of 'rigid sphere' equation which describes helium state in bubbles

  14. Helium desorption in EFDA iron materials for use in nuclear fusion reactors; Desorcion de helio en materiales de fierro EFDA para su aplicacion en los reactores de fusion nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Salazar R, A. R.; Pinedo V, J. L. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sanchez, F. J.; Ibarra, A.; Vila, R., E-mail: arsr2707@hotmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense No. 40, 28040 Madrid (Spain)

    2015-09-15

    In this paper the implantation with monoenergetic ions (He{sup +}) was realized with an energy of 5 KeV in iron samples (99.9999 %) EFDA (European Fusion Development Agreement) using a collimated beam, after this a Thermal Desorption Spectrometry of Helium (THeDS) was made using a leak meter that detects amounts of helium of up to 10{sup -}- {sup 12} mbar l/s. Doses with which the implantation was carried out were 2 x 10{sup 15} He{sup +} /cm{sup 2}, 1 x 10{sup 16} He{sup +} /cm{sup 2}, 2 x 10{sup 16} He{sup +} /cm{sup 2}, 1 x 10{sup 17} He{sup +} /cm{sup 2} during times of 90 s, 450 s, 900 s and 4500 s, respectively. Also, using the SRIM program was calculated the depth at which the helium ions penetrate the sample of pure ion, finding that the maximum distance is 0.025μm in the sample. For this study, 11 samples of Fe EFDA were prepared to find defects that are caused after implantation of helium in order to provide valuable information to the manufacture of materials for future fusion reactors. However understand the effects of helium in the micro structural evolution and mechanical properties of structural materials are some of the most difficult questions to answer in materials research for nuclear fusion. When analyzing the spectra of THeDS was found that five different groups of desorption peaks existed, which are attributed to defects of He caused in the material, these defects are He{sub n} V (2≤n≤6), He{sub n} V{sub m}, He V for the groups I, II and IV respectively. These results are due to the comparison of the peaks presented in the desorption spectrum of He, with those of other authors who have made theoretical calculations. Is important to note that the thermal desorption spectrum of helium was different depending on the dose with which the implantation of He{sup +} was performed. (Author)

  15. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  16. Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10

    Directory of Open Access Journals (Sweden)

    Chengxiang Guo

    2016-01-01

    Full Text Available Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR. To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10 in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for γ dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

  17. Liquid helium

    CERN Document Server

    Atkins, K R

    1959-01-01

    Originally published in 1959 as part of the Cambridge Monographs on Physics series, this book addresses liquid helium from the dual perspectives of statistical mechanics and hydrodynamics. Atkins looks at both Helium Three and Helium Four, as well as the properties of a combination of the two isotopes. This book will be of value to anyone with an interest in the history of science and the study of one of the universe's most fundamental elements.

  18. Helium behaviour in nuclear glasses

    International Nuclear Information System (INIS)

    Fares, T.

    2011-01-01

    The present thesis focuses on the study of helium behavior in R7T7 nuclear waste glass. Helium is generated by the minor actinides alpha decays incorporated in the glass matrix. Therefore, four types of materials were used in this work. These are non radioactive R7T7 glasses saturated with helium under pressure, glasses implanted with 3 He + ions, glasses doped with curium and glasses irradiated in nuclear reactor. The study of helium solubility in saturated R7T7 glass has shown that helium atoms are inserted in the glass free volume. The results yielded a solubility of about 10 16 at. cm -3 atm. -1 . The incorporation limit of helium in this type of glass has been determined; its value amounted to about 2*10 21 at. cm -3 , corresponding to 2.5 at.%. Diffusion studies have shown that the helium migration is controlled by the single population dissolved in the glass free volume. An ideal diffusion model was used to simulate the helium release data which allowed to determine diffusion coefficients obeying to the following Arrhenius law: D = D 0 exp(-E a /kBT), where D 0 = 2.2*10 -2 and 5.4*10 -3 cm 2 s -1 and E a = 0.61 eV for the helium saturated and the curium doped glass respectively. These results reflect a thermally activated diffusion mechanism which seems to be not influenced by the glass radiation damage and helium concentrations studied in the present work (up to 8*10 19 at. g -1 , corresponding to 0.1 at.%). Characterizations of the macroscopic, structural and microstructural properties of glasses irradiated in nuclear reactor did not reveal any impact associated with the presence of helium at high concentrations. The observed modifications i.e. a swelling of 0.7 %, a decrease in hardness by 38 %, an increase between 8 and 34 % of the fracture toughness and a stabilization of the glass structure under irradiation, were attributed to the glass nuclear damage induced by the irradiation in reactor. Characterizations by SEM and TEM of R7T7 glasses implanted

  19. Detail analysis of tritium permeation in the metal liquid channels of the regenerating sheaths of a fusion reactor in presence of helium bubbles; Analisis de detalle de la permeacion de tritio en los caneles de metal liquido de las envolturas regeneradoras de un reactor de fusion en presencia de burbujas de helio

    Energy Technology Data Exchange (ETDEWEB)

    Banet, L.; Mas de les Valls, E.; Sedano, L. A.

    2012-07-01

    Inside the channels of liquid metal of the fusion reactor regenerative wrappers, the possible existence of nucleated helium bubbles is not remote. Helium is formed joined the tritium in the escaped neutrons of plasma with lithium. The accumulation of helium in the contact surfaces, between the structure and ML, lead a reduction of heat transfer, at the same time a reduction in the permeation of tritium. The coexistence of three phases in touch: metal liquid, helium and structural material, makes the transport of heat and tritium in a complex phenomenon. To enrich tritium transport studies conducted in the past, there is now a detail analysis of the helium bubble environment adhered to the channel ML wall of a regenerative wrap. For the study we used a CFD tool development on free code OpenFOAM.

  20. Helium cryogenics

    CERN Document Server

    Van Sciver, Steven W

    2012-01-01

    Twenty five years have elapsed since the original publication of Helium Cryogenics. During this time, a considerable amount of research and development involving helium fluids has been carried out culminating in several large-scale projects. Furthermore, the field has matured through these efforts so that there is now a broad engineering base to assist the development of future projects. Helium Cryogenics, 2nd edition brings these advances in helium cryogenics together in an updated form. As in the original edition, the author's approach is to survey the field of cryogenics with emphasis on helium fluids. This approach is more specialized and fundamental than that contained in other cryogenics books, which treat the associated range of cryogenic fluids. As a result, the level of treatment is more advanced and assumes a certain knowledge of fundamental engineering and physics principles, including some quantum mechanics. The goal throughout the work is to bridge the gap between the physics and engineering aspe...

  1. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Bjornard, Trond; Hockert, John

    2011-01-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

  2. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials

    International Nuclear Information System (INIS)

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H 2 /5 Pa CO/5 Pa CH 4 / 2 O (500 μatm H 2 /50 μatm CO/50 μatm CH 4 / 2 O) at 900 0 C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables

  3. Variable electricity and steam from salt, helium and sodium cooled base-load reactors with gas turbines and heat storage - 15115

    International Nuclear Information System (INIS)

    Forsberg, C.; McDaniel, P.; Zohuri, B.

    2015-01-01

    Advances in utility natural-gas-fired air-Brayton combed cycle technology is creating the option of coupling salt-, helium-, and sodium-cooled nuclear reactors to Nuclear air-Brayton Combined Cycle (NACC) power systems. NACC may enable a zero-carbon electricity grid and improve nuclear power economics by enabling variable electricity output with base-load nuclear reactor operations. Variable electricity output enables selling more electricity at times of high prices that increases plant revenue. Peak power is achieved using stored heat or auxiliary fuel (natural gas, bio-fuels, hydrogen). A typical NACC cycle includes air compression, heating compressed air using nuclear heat and a heat exchanger, sending air through a turbine to produce electricity, reheating compressed air, sending air through a second turbine, and exhausting to a heat recovery steam generator (HRSG). In the HRSG, warm air produces steam that is used to produce added electricity. For peak power production, auxiliary heat (natural gas, stored heat) is added before the air enters the second turbine to raise air temperatures and power output. Like all combined cycle plants, water cooling requirements are dramatically reduced relative to other power cycles because much of the heat rejection is in the form of hot air. (authors)

  4. Bed system performance in helium circulation mode

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yean Jin; Jung, Kwang Jin; Ahn, Do Hee; Chung, Hong Suk [UST, Daejeon (Korea, Republic of); Kang, Hee Suk [KAERI, Daejeon (Korea, Republic of); Yun, Sei Hun [NFRI, Deajeon (Korea, Republic of)

    2016-05-15

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, We have conducted an experiment for storing hydrogen to depleted uranium and zirconium cobalt. The helium blanket effect has been observed in experiments using metal hydrides. The collapse of the hydrogen isotopes are accompanied by the decay heat and helium-3. Helium-3 dramatically reduces the hydrogen isotope storage capacity by surrounding the metal. This phenomenon is called a helium blanket effect. In addition the authors are working on the recovery and removal techniques of helium-3. In this paper, we discuss the equipment used to test the helium blanket effect and the results of a helium circulation experiment. The helium-3 produced surrounds the storage material surface and thus disturbs the reaction of the storage material and the hydrogen isotope. Even if the amount of helium-3 is small, the storage capacity of the SDS bed significantly drops. This phenomenon is the helium blanket effect. To resolve this phenomenon, a circulating loop was introduced. Using a circulating system, helium can be separated from the storage material. We made a helium loop that includes a ZrCo bed. Then using a metal bellows pump, we tested the helium circulation.

  5. Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1981-01-15

    The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

  6. Checking the sealing of fuel elements by helium sweating - case of the reactors G2 (1960); Controle de l'etancheite des elements combustibles par ressuage d'helium - cas du reacteur G2 (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Blanc, B; D' Orival, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Choumoff, S [Compagnie Francaise Thomson-Houston, 75 - Paris (France)

    1960-07-01

    The G2 slug is a welded, hermetically sealed unit; the seal is checked by placing the fuel element in a helium atmosphere under pressure, then measuring the quantity of helium it releases in a vessel under vacuum. The theoretical aspect and the conditions of industrial application are reviewed, and the installations described. (author) [French] La cartouche G2 se presente comme un ensemble soude, hermetique; le controle d'etancheite s'effectue en immergeant l'element combustible dans une atmosphere d'helium sous pression puis en mesurant la quantite d'helium qu'il restitue dans une enceinte sous vide. L'aspect theorique et les conditions d'exploitation industrielle sont evoques et les installations decrites. (auteur)

  7. Interactions of reactor helium and simulating gas mixtures with high-temperature metals with particular regard to simultaneous deformation

    International Nuclear Information System (INIS)

    Berchtold, L.

    1983-01-01

    For the observation of multicomponent alloys (Inconel 617 and 713LC, chroman (Ni20Cr), vacromium (Ni20Cr+Si), TZM) in multicomponent HTR atmospheres (HHT search gas), interaction between gases and metals was studied, both in theoretical descriptions and experimentally. From the experimental viewpoint, gradual simplification employs, on the one hand, tests effected in undiluted atmospheres with exclusively oxidizing or carburizing properties; on the other hand, more simple alloys and pure metals are applied specifically in the helium atmosphere. For an evaluation of the materials, it is maintained that in a strongly oxidizing (H 2 O-rich) atmosphere, e.g. in HHT search gas, materials with sufficient chrome content (e.g. 20% Cr in Ni alloys such as IN 617) offer favourable conditions for an almost complete interruption of carburizing reactions. In that case, the maintenance of the shielding effect of coating during rapid deformation and a tendency to planar delamination during deformation, which becomes stronger as the layer thickness increases, appear to be critical. Concentrations of oxide-forming agents stronger than chromium offer disadvantages rather than advantages. Owing to its tendency to flake off as the covering oxide SiO 2 or as part of a cover layer, silicon may more than destroy the light advantage of a slowed down process of carbon diffusion. The cast alloy IN 713LC shows a deep-reaching carburation in HHT search gas, both with and without deformation. No deep-reaching corrosive damage is noticeable on the molybdenum alloy TZM. (orig./MM) [de

  8. Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

    Directory of Open Access Journals (Sweden)

    Nunnenmann Elena

    2017-01-01

    Full Text Available An uncertainty analysis was performed for the tritium breeding ratio (TBR of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB, currently under development in the European Power Plant Physics and Technology (PPPT programme for a fusion power demonstration reactor (DEMO. A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

  9. Standard Guide for Simulation of Helium Effects in Irradiated Metals

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This guide provides advice for conducting experiments to investigate the effects of helium on the properties of metals where the technique for introducing the helium differs in some way from the actual mechanism of introduction of helium in service. Simulation techniques considered for introducing helium shall include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended. 1.2 Two other methods for introducing helium into irradiated materials are not covered in this guide. They are the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, and isotopic tailoring in both fast and mixed-spectrum fission reactors. These techniques are described in Refs (1-5). Dual ion beam techniques (6) for simultaneously implanting helium and generating displacement damage are also not included here. This lat...

  10. Creep properties of base metal and welded joint of Hastelloy XR produced for High-Temperature Engineering Test Reactor in simulated primary coolant helium

    International Nuclear Information System (INIS)

    Kurata, Yuji; Tsuji, Hirokazu; Shindo, Masami; Suzuki, Tomio; Tanabe, Tatsuhiko; Mutoh, Isao; Hiraga, Kenjiro

    1999-01-01

    Creep tests of base metal, weld metal and welded joint of Hastelloy XR, which had the same chemical composition as Hastelloy XR produced for an intermediate heat exchanger of the High-Temperature Engineering Test Reactor, were conducted in simulated primary coolant helium. The weld metal and welded joint showed almost equal to or longer rupture time than the base metal of Hastelloy XR at 850 and 900degC, although they gave shorter rupture time at 950degC under low stress and at 1,000degC. The welded joint of Hastelloy XR ruptured at the base metal region at 850 and 900degC. On the other hand, it ruptured at the weld metal region at 950 and 1,000degC. The steady-state creep rate of weld metal of Hastelloy XR was lower than that of base metal at 850, 900 and 950degC. The creep rupture strengths of base metal, weld metal and welded joint of Hastelloy XR obtained in this study were confirmed to be much higher than the design allowable creep-rupture stress (S R ) of the Design Allowable Limits below 950degC. (author)

  11. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico

    International Nuclear Information System (INIS)

    Barragan M, A.M.; Martin del Campo M, C.

    2007-01-01

    A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

  12. Analysis of multi-scale spatial separation in a block-type thorium-loaded helium-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Huang, Jie; Ding, Ming

    2017-01-01

    Highlights: • Four-level of spatial separation is described in a block-type thorium-loaded HTR. • A traditional two-step calculation scheme is used to get the neutronic performance. • Fuel cycle cost is calculated by the levelised lifetime cost method. • Fuel cycle cost decreases with the increase of separation level or thorium content. • Effective enrichment basically determines the fuel cycle cost. - Abstract: With nuclear energy’s rapid development in recent years, supply of nuclear fuel has become increasingly important. Thorium has re-gained attention because of its abundant reserves and excellent physical properties. Compared to the homogeneous Th/U MOX fuel, separation of thorium and uranium in space is a better use of thorium. Therefore, this paper describes four-level spatial separation – no separation, tristructural-isotropic (TRISO) level, channel level and block level – in a block-type thorium-loaded helium-cooled high-temperature reactor (HTR). A traditional two-step calculation scheme, lattice calculation followed by core calculation, is used to get the neutronic performance of the equilibrium cycle, including uranium enrichment, mass of fuel, effective multiplication factor, and average conversion ratio. Based on these data, the fuel cycle cost of different-scale spatial separation can be calculated by the levelised lifetime cost method as a function of thorium content. As the separation level increases from no separation to channel level, the effective enrichment decreases 15% due to the increase of resonance escape probability. So there is a 13% drop for the fuel cycle cost. For TRISO-level separation, as the thorium content increases from 9 to 57%, the effective enrichment decreases 14% because of the superior breeding capacity of U-233. As a result, the fuel cycle cost also has about a 12% decrease. From the perspective of fuel cycle economics, channel-level separation with 60% thorium content is suggested.

  13. Neutral helium beam probe

    Science.gov (United States)

    Karim, Rezwanul

    1999-10-01

    This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

  14. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico; Estudio de tecnicas de agregacion de indicadores para la seleccion de un nuevo reactor nuclear para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.M.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, 04510 Mexico D.F. (Mexico)]. e-mail: ale_bar_m@yahoo.com.mx

    2007-07-01

    A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

  15. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  16. Helium crystals

    International Nuclear Information System (INIS)

    Lipson, S.G.

    1987-01-01

    Hexagonal close-packed helium crystals in equilibrium with superfluid have been found to be one of the few systems in which an anisotropic solid comes into true thermodynamic equilibrium with its melt. The discovery of roughening transitions at the liquid-solid interface have shown this system to be ideal for the study of the statistical mechanics of interface structures. We describe the effect of roughening on the shape and growth of macroscopic crystals from both the theoretical and experimental points of view. (author)

  17. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor

    International Nuclear Information System (INIS)

    Tauveron, N.

    2006-02-01

    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge

  18. The installation of helium auxiliary systems in HTGR

    International Nuclear Information System (INIS)

    Qin Zhenya; Fu Xiaodong

    1993-01-01

    The inert gas Helium was chosen as reactor coolant in high temperature gas coolant reactor, therefore a set of Special and uncomplex helium auxiliary systems will be installed, the safe operation of HTR-10 can be safeguarded. It does not effect the inherent safety of HTR-10 MW if any one of all those systems were damaged during operation condition. This article introduces the design function and the system principle of all helium auxiliary systems to be installed in HTR-10. Those systems include: helium purification and its regeneration system, helium supply and storage system, pressure control and release system of primary system, dump system for helium auxiliary system and fuel handling, gaseous waste storage system, water extraction system for helium auxiliary systems and evacuation system for primary system

  19. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core; Comportamento termofluidodinamico do gas refrigerante helio em um canal topico de reator VHTGR de nucleo prismatico

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Allan Cavalcante

    2016-08-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4{sup th} generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range

  20. Development of monitoring system using acoustic emission for detection of helium gas leakage for primary cooling system and flow-induced vibration for heat transfer tube of heat exchangers for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Kunitomi, Kazuhiko; Furusawa, Takayuki; Shinozaki, Masayuki; Satoh, Yoshiyuki; Yanagibashi, Minoru

    1998-10-01

    The High Temperature Engineering Test Reactor (HTTR) uses helium gas for its primary coolant, whose leakage inside reactor containment vessel is considered in design of the HTTR. It is necessary to detect leakage of helium gas at an early stage so that total amount of the leakage should be as small as possible. On the other hand, heat transfer tubes of heat exchangers of the HTTR are designed not to vibrate at normal operation, but the flow-induced vibration is to be monitored to provide against an emergency. Thus monitoring system of acoustic emission for detection of primary coolant leakage and vibration of heat transfer tubes was developed and applied to the HTTR. Before the application to the HTTR, leakage detection test was performed using 1/4 scaled model of outer tube of primary concentric hot gas duct. Result of the test covers detectable minimum leakage rate and effect of difference in gas, pressure, shape of leakage path and distance from the leaking point. Detectable minimum leakage rate was about 5 Ncc/sec. The monitoring system is promising in leakage detection, though countermeasure to noise is to be needed after the HTTR starts operating. (author)

  1. Superfluid helium at subcritical active core

    International Nuclear Information System (INIS)

    Vasil'ev, V.V.; Lopatkin, A.V.; Muratov, V.G.; Rakhno, I.L.

    2002-01-01

    Power range and neutron flux wherein super thermal source was realized at high volume of superfluid helium were investigated. MCU, BRAND, MCNP codes were used for the calculation of reactors. It is shown that the availability of full-size diameter for cryogenic source of ultracold neutrons, as the source with superfluid helium is considered, is possible in the reflector of subcritical assembly. Results obtained from the MCNP-4B code application demonstrated that the density of thermal neutron flux in helium must be not higher than 2.3 x 10 11 s -1 cm -2 [ru

  2. Study on restriction method for end-wall boundary layer thickness in axial helium gas compressor for gas turbine high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kunitomi, Kazuhiko; Inagaki, Yoshiyuki

    2009-01-01

    Aerodynamic performance test was carried out using a 1/3 scale, 4-stage model of the helium gas compressor to investigate an effect of end-wall over-camber to prevent decrease of axial velocity in the end-wall boundary layer. The model compressor consists of a rotor, 500 mm in diameter, which is driven by an electric motor at a rotational speed of 10800 rpm. The rotor blade span of the first stage is 34 mm. The test was carried out under the condition that the helium gas pressure of 0.88 MPa, temperature of 30degC, and mass flow rate of 12.47 kg/s at the inlet. A 3-dimensional aerodynamic code, which was verified using the test data, showed that axial velocity was lowered by using a blade which increased the inlet blade angle around the end-wall region of the casing side in comparison with that using the original design blade, because the inlet flow angle mismatched with the inlet blade angle of the rotor blade, as opposed to the prediction by a conventional air compressor design method. The overall adiabatic efficiency of the full scale 20-stage helium gas compressor was predicted 89.7% from the Reynolds number dependency of the test data by using the original design blade. (author)

  3. Liquid helium target

    International Nuclear Information System (INIS)

    Fujii, Y.; Kitami, T.; Torikoshi, M.

    1984-12-01

    A liquid helium target system has been built and used for the experiment on the reaction 4 He(γ, p). The target system has worked satisfactorily; the consumption rate of liquid helium is 360 ml/h and the cryogenic system retains liquid helium for about ten hours. The structure, operation and performance of the target system are reported. (author)

  4. Simulation study of radiation damage induced by energetic helium nuclei

    International Nuclear Information System (INIS)

    Hoang Dac Luc; Vo Tuong Hanh; Hoang Dac Dat

    2003-01-01

    High energy alpha particles produced by neutron-induced nuclear reactions can damage severely reactor materials. Simulation of this process is described using theoretical calculation and ion irradiation experiments at different displacement doses and Helium doses. (author)

  5. Simulation study of radiation damage induced by energetic helium nuclei

    CERN Document Server

    Hoang Dac Luc; Hoang Dac Dat

    2003-01-01

    High energy alpha particles produced by neutron-induced nuclear reactions can damage severely reactor materials. Simulation of this process is described using theoretical calculation and ion irradiation experiments at different displacement doses and Helium doses.

  6. Effect of carburizing helium environment on creep behavior of Ni-base heat-resistant alloys for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kurata, Yuji; Ogawa, Yutaka; Nakajima, Hajime

    1988-01-01

    Creep tests were conducted on Ni-base heat-resistant alloys Hastelloy XR and XR-II, i.e. versions of Hastelloy X modified for nuclear applications, at 950degC using four types of helium environment with different impurity compositions, and mainly the effect of carburization was examined. For all the materials tested, the values of creep rupture time obtained under the carburizing conditions were similar to or longer than those in the commonly used, standard test environment (JAERI Type B helium). The difference among the results was interpreted by the counterbalancing effects of the strengthening due to carburization and possible weakening caused under very low oxidizing potential. In the corrosion monitoring specimens pronounced carbon pick-up was observed in the environment with high carbon activity and very low oxidizing potential. Based on the results obtained in the present and the previous works, it is suggested that a moderate control of the impurity chemistry is important rather than simple purification of the coolant in protecting the material from the environment-enhanced degradation. Either condition with high or low extremes in the oxidizing and carburizing potentials may cause enhanced degradation and thus are desirable to be avoided at the elevated temperatures. (author)

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  8. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  9. Helium storage and control system for the PBMR

    International Nuclear Information System (INIS)

    Verkerk, E.C.

    1997-01-01

    The power conversion unit will convert the heat energy in the reactor core to electrical power. The direct-closed cycle recuperated Brayton Cycle employed for this concept consists of a primary helium cycle with helium powered turbo compressors and a power turbine. The helium is actively cooled with water before the compression stages. A recuperator is used to preheat the helium before entering the core. The start of the direct cycle is initiated by a mass flow from the helium inventory and control system via a jet pump. When the PBMR is connected to the grid, changes in power demand can be followed by changing the helium flow and pressure inside the primary loop. Small rapid adjustments can be performed without changing the helium inventory of the primary loop. The stator blade settings on the turbines and compressors are adjustable and it is possible to bypass reactor and turbine. This temporarily reduces the efficiency at which the power conversion unit is operating. Larger or long term adjustments require storage or addition of helium in order to maintain a sufficient level of efficiency in the power conversion unit. The helium will be temporarily stored in high pressure tanks. After a rise in power demand it will be injected back into the system. Some possibilities how to store the helium are presented in this paper. The change of helium inventory will cause transients in the primary helium loop in order to acquire the desired power level. At this stage, it seems that the change of helium inventory does not strongly effect the stability of the power conversion unit. (author)

  10. Corrosion of graphitic high temperature reactor materials in steam/helium mixtures at total pessures of 3-55 bar and temperatures of 900-1150 C (1173-1423K)

    International Nuclear Information System (INIS)

    Hinssen, H.K.; Loenissen, K.J.; Katscher, W.; Moormann, R.

    1993-03-01

    In course of accident examination for (HTR), experiments on the corrosion behavior of graphitic reactor materials in steam have been performed a total pressures of 3-55bar and temperatures of 900-1150 C (1173-1423K); these experiments and their evaluation are documented here. Reactor materials examined are the structure graphite V483T2 and the fuel element matrices A3-27 and A3-3. In all experiments, the steam partial pressure was 474mbar (inert gas helium). The dependence of reaction rates and density profiles on burn-off, total pressure and temperature has been examined. Experimental reaction rates depending on burn-off are fitted by theoretical curves, a procedure, which allows rate comparison for a well defined burn-off. Comparing rates as a function of total pressure, V483T2 shows a linear dependence on 1√p total , whereas for matrix materials a pressure independent rate was found for p total 4mm for A3-3. (orig.) [de

  11. Design of multi-input multi-output controller for magnetic bearing which suspends helium gas-turbine generator rotor for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Takada, Shoji; Funatake, Yoshio; Inagaki, Yoshiyuki

    2009-01-01

    A design of a MIMO controller, which links magnetic forces of multiple magnetic bearings by feedback of multiple measurement values of vibration of a rotor, was proposed for the radial magnetic bearings for the generator rotor of helium gas turbine with a power output of 300 MWe. The generator rotor is a flexible rotor, which passes over the forth critical speed. A controller transfer function was derived at the forth critical speed, in which the bending vibration mode is similar to the one which is excited by unbalance mass to reduce a modeling error. A 1404-dimensional un-symmetric coefficient matrix of equation of state for the rotating rotor affected by Jayro effect was reduced by a modal decomposition using Schur decomposition to reduce a reduction error. The numerical results showed that unbalance response of rotor was 53 and 80 μm p-p , respectively, well below the allowable limits both at the rated and critical speeds. (author)

  12. Helium leak and chemical impurities control technology in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Shimizu, Atsushi; Hamamoto, Shimpei; Sakaba, Nariaki

    2014-01-01

    Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown. (author)

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  15. Neutron-induced helium implantation in GCFR cladding

    International Nuclear Information System (INIS)

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  16. Evaluation of effect of inlet distortion on aerodynamic performance of helium gas compressor for gas turbine high temperature reactor (GTHTR300). Contract research

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kurokouchi, Naohiro; Kunitomi, Kazuhiko

    2006-02-01

    Because the main pipe is connected perpendicular to the flow direction inside the distributing header in the inlet casing of the helium gas compressor design of GTHTR300, the main flow flowing into the header tends to separate from the header wall and to cause reverse flow, which increases flow resistance in the header. This phenomenon increases the total pressure loss in the header and inlet distortion, which is considered to deteriorate the aerodynamic performance of the compressor. Tests were carried out to evaluate the effects of inlet distortion on aerodynamic performance of compressor by using a 1/3-scale helium gas compressor model by varying a level of inlet distortion. Flow was injected from the wall of header to make circumferential velocities uniform before and after the reverse flow region to dissipate the separation and reverse flow. At the design point, inlet distortion was reduced by 2-3% by injection, which resulted in increasing adiabatic efficiency of blade section by 0.5%. A modified flow rate at surge point was lowered from 10.0 kg/s to 9.6 kg/s. At the same time, pressure loss of the inlet casing was reduced by 3-5 kPa, which is equivalent to adiabatic efficiency improvement around 0.8%. By setting orifice at the inlet of the inlet casing, the level of inlet distortion became 3% higher and the adiabatic efficiency of blade section became 1% higher at the design point. The modified flow rate at surge point increased from 10.6 to 10.9 kg/s. A new correlation between inlet distortion and adiabatic efficiency of blade section at the rated flow rate was derived based on compressor-in-parallel model and fitted to the test results. An overall adiabatic efficiency of full-scale compressor was predicted 90.2% based on the test results of efficiency and Reynolds number correlation, which was close to 89.7% that was predicted by test calibrated design through-flow code. (author)

  17. State of the Art Report for a Bearing for VHTR Helium Circulator

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Song, Kee Nam; Kim, Yong Wan; Lee, Won Jae

    2008-10-01

    A helium circulator in a VHTR(Very High Temperature gas-cooled Reactor) plays a core role which translates thermal energy at high temperature from a nuclear core to a steam generator. Helium as a operating coolant circulates a primary circuit in high temperature and high pressure state, and controls thermal output of a nuclear core by controlling flow rate. A helium circulator is the only rotating machinery in a VHTR, and its reliability should be guaranteed for reliable operation of a reactor and stable production of hydrogen. Generally a main helium circulator is installed on the top of a steam generator vessel, and helium is circulated only by a main helium circulator in a normal operation state. An auxiliary or shutdown circulator is installed at the bottom of a reactor vessel, and it is an auxiliary circulator for shutting down a reactor in case of refueling or accelerating cooling down in case of fast cooling. Since a rotating shaft of a helium circulator is supported by bearings, bearings are the important machine elements which determines reliability of a helium circulator and a nuclear reactor. Various types of support bearings have been developed and applied for circulator bearings since 1960s, and it is still developing for developing VHTRs. So it is necessary to review and analyze the current technical state of helium circulator support bearings to develop bearings for Koran developing VHTR helium circulator

  18. Cooling with Superfluid Helium

    Energy Technology Data Exchange (ETDEWEB)

    Lebrun, P; Tavian, L [European Organization for Nuclear Research, Geneva (Switzerland)

    2014-07-01

    The technical properties of helium II (‘superfluid’ helium) are presented in view of its applications to the cooling of superconducting devices, particularly in particle accelerators. Cooling schemes are discussed in terms of heat transfer performance and limitations. Large-capacity refrigeration techniques below 2 K are reviewed, with regard to thermodynamic cycles as well as process machinery. Examples drawn from existing or planned projects illustrate the presentation. Keywords: superfluid helium, cryogenics.

  19. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  20. Helium the disappearing element

    CERN Document Server

    Sears, Wheeler M

    2015-01-01

    The subject of the book is helium, the element, and its use in myriad applications including MRI machines, particle accelerators, space telescopes, and of course balloons and blimps. It was at the birth of our Universe, or the Big Bang, where the majority of cosmic helium was created; and stellar helium production continues. Although helium is the second most abundant element in the Universe, it is actually quite rare here on Earth and only exists because of radioactive elements deep within the Earth. This book includes a detailed history of the discovery of helium, of the commercial industry built around it, how the helium we actually encounter is produced within the Earth, and the state of the helium industry today. The gas that most people associate with birthday party balloons is running out. “Who cares?” you might ask. Well, without helium, MRI machines could not function, rockets could not go into space, particle accelerators such as those used by CERN could not operate, fiber optic cables would not...

  1. Helium dilution refrigerator

    International Nuclear Information System (INIS)

    1973-01-01

    A new system of continuous heat exchange for a helium dilution refrigerator is proposed. The 3 He effluent tube is concurrent with the affluent mixed helium tube in a vertical downward direction. Heat exchange efficiency is enhanced by placing in series a number of elements with an enlarged surface area

  2. Helium localisation in tritides

    International Nuclear Information System (INIS)

    Flament, J.L.; Lozes, G.

    1982-06-01

    Study of titanium and LaNi 5 type alloys tritides lattice parameters evolution revealed that helium created by tritium decay remains in interstitial sites up to a limit material dependant concentration. Beyond this one exceeding helium precipites in voids [fr

  3. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  4. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  5. Study on the conversion of H2 and CO from the helium carrier gas of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liao Cuiping; Zheng Zhenhong; Shi Fuen

    1995-01-01

    The conversions of hydrogen and carbon monoxide into water vapor and carbon dioxide on CuO-ZnO-Al 2 O 3 catalyst are studied. The effects of different temperature, system atmospheric pressure, impurity gas concentration, flow and dew point on properties of cupric oxide bed are investigated. The conversion characteristics curves of H 2 and CO are given. Experimental data of conversion capacity, action period and conversion efficiency of CuO-ZnO-Al 2 O 3 are obtained and the optimal parameters are determined. The results show that the concentration of H 2 and CO of the effluent gas after purification can reach below 2 x 10 -6 , respectively. So it can meet the demands of high temperature gas-cooled reactor and also provide optimal design parameters and reliable data for conversion of H 2 and CO on CuO-ZnO-Al 2 O 3 catalyst

  6. A liquid helium saver

    International Nuclear Information System (INIS)

    Avenel, O.; Der Nigohossian, G.; Roubeau, P.

    1976-01-01

    A cryostat equipped with a 'liquid helium saver' is described. A mass flow rate M of helium gas at high pressure is injected in a counter-flow heat exchanger extending from room to liquid helium temperature. After isenthalpic expansion through a calibrated flow impedance this helium gas returns via the low pressure side of the heat exchanger. The helium boil-off of the cryostat represents a mass flow rate m, which provides additional precooling of the incoming helium gas. Two operating regimes appear possible giving nearly the same efficiency: (1) high pressure (20 to 25 atm) and minimum flow (M . L/W approximately = 1.5) which would be used in an open circuit with helium taken from a high pressure cylinder; and (2) low pressure (approximately = 3 atm), high flow (M . L/W > 10) which would be used in a closed circuit with a rubber diaphragm pumping-compressing unit; both provide a minimum theoretical boil-off factor of about 8%. Experimental results are reported. (U.K.)

  7. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  8. Hot helium flow test facility summary report

    International Nuclear Information System (INIS)

    1980-06-01

    This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

  9. Tritium Decay Helium-3 Effects in Tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Merrill, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructural evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium

  10. The evolution of US helium-cooled blankets

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

    1991-01-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

  11. Structural changes in a copper alloy due to helium implantation

    International Nuclear Information System (INIS)

    Moreno, D.; Eliezer, D.

    1996-01-01

    The most suitable nuclear fusion reaction for energy production occurs between the two heavy hydrogen isotopes, deuterium and tritium. Structural materials in fusion reactors will be exposed to helium implantation over a broad range of energies. The deformation and partial exfoliation of surface layers due to hydrogen isotopes and helium contribute to the total erosion of the first wall. For this reason, one of the most important criteria in the choice of materials for the first wall of fusion reactors is the material's damage resistance. Recent advances in developing nuclear fusion reactors reveal that efficient heat removal from plasma-facing components is very important. Copper and copper alloys are considered an attractive choice for transporting such a high heat flux without thermal damage as they have high thermal conductivity. In the present study the authors report on the structural changes in a copper alloy, due to the helium implantation on the very near surface area, observed by transmission electron microscopy

  12. Evaluation of helium cooling for fusion divertors

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

  13. Experimental method for investigating helium effects in irradiated vanadium

    International Nuclear Information System (INIS)

    Smith, D.L.; Matsui, H.; Greenwood, L.; Loomis, B.

    1987-10-01

    Analyses have been performed which indicate that an effective method for experimentally investigating helium effects in neutron irradiated vanadium base alloys can be developed. The experimental procedure involves only modest modifications to existing procedures currently used for irradiation testing of vanadium-base alloys in the FFTF reactor. Helium is generated in the vanadium alloy by decay of tritium which is either preinjected or generated within the test capsule. Calculations indicate that nearly constant He/dpa ratios of desired magnitude can be attained by proper selection of experimental parameters. The proposed method could have a major impact on the development of vanadium base alloys for fusion reactor applications. 8 refs., 4 figs

  14. High-temperature helium embrittlement (T>=0,45Tsub(M)) of metals

    International Nuclear Information System (INIS)

    Batfalsky, P.

    1984-06-01

    High temperature helium embrittlement, swelling and irradiation creep are the main technical problem of fusion reactor materials. The expected helium production will be very high. The helium produced by (n,α)-processes precipitates into helium bubbles because its solubility in solid metals is very low. Under continuous helium production at high temperature and stress the helium bubbles grow and lead to intergranular early failure. Solution annealed foil specimens of austenitic stainless steel AISI 316 were implanted with α-particles: 1. during creep tests at 1023 K (''in-beam'' test) 2. before the creep tests at high temperature (1023 K). The creep tests have been performed within large ranges of test parameter, e.g. applied stress, temperature, helium implantation rate and helium concentration. After the creep tests the microstructure was investigated using scanning (SEM) and transmission (TEM) electron microscopy. All the helium implanted specimens showed high temperature helium embrittlement, i.e. reduction of rupture time tsub(R) and ductility epsilonsub(R) and evidence of intergranular brittle fracture. The ''in-beam'' creep tests showed greater reduction of rupture time tsub(R) and ductility than the preimplanted creep tests. The comparison of this experimentally obtained data with various theoretical models of high temperature helium embrittlement showed that within the investigated parameter ranges the mechanism controlling the life time of the samples is probably the gas driven stable growth of the helium bubbles within the grain boundaries. (orig.)

  15. Conceptual design of helium experimental loop

    International Nuclear Information System (INIS)

    Yu Xingfu; Feng Kaiming

    2007-01-01

    In a future demonstration fusion power station (DEMO), helium is envisaged as coolant for plasma facing components, such as blanket and dive,or. All these components have a very complex geometry, with many parallel cooling channels, involving a complex helium flow distribution. Test blanket modules (TBM) of this concept will under go various tests in the experimental reactor ITER. For the qualification of TBM, it is indispensable to test mock-ups in a helium loop under realistic pressure and temperature profiles, in order to validate design codes, especially regarding mass flow and heat transition processes in narrow cooling channels. Similar testing must be performed for DEMO blanket, currently under development. A Helium Experimental Loop (HELOOP) is planed to be built for TBM tests. The design parameter of temperature, pressure, flow rate is 550 degree C, 10 MPa, l kg/s respectively. In particular, HELOOP is able to: perform full-scale tests of TBM under realistic conditions; test other components of the He-cooling system in ITER; qualify the purification circuit; obtain information for the design of the ITER cooling system. The main requirements and characteristics of the HELOOP facility and a preliminary conceptual design are described in the paper. (authors)

  16. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor; Simulation numerique et analyse du declenchement et du developpement des instabilites axiales dans les turbomachines: application a un transitoire de breche dans un reacteur nucleaire a helium

    Energy Technology Data Exchange (ETDEWEB)

    Tauveron, N

    2006-02-15

    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge

  17. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor; Simulation numerique et analyse du declenchement et du developpement des instabilites axiales dans les turbomachines: application a un transitoire de breche dans un reacteur nucleaire a helium

    Energy Technology Data Exchange (ETDEWEB)

    Tauveron, N

    2006-02-15

    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge

  18. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  19. Design and study of Engineering Test Facility - Helium Circulator

    International Nuclear Information System (INIS)

    Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

    2015-01-01

    Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

  20. Ultralow temperature helium compressor for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Asakura, Hiroshi

    1988-01-01

    Ishikawajima Harima Heavy Industries Co., Ltd. started the development of an ultralow temperature helium compressor for helium liquefaction in 1984 jointly with Japan Atomic Energy Research Institute, and has delivered the first practical machine to the Superconductive Magnet Laboratory of JAERI. For a large superconductive magnet to be used in the stable state for a fusion reactor, conventional superconductive materials (NbTi, NbTi 3 Sn, etc.) must be used, being cooled forcibly with supercritical helium. The supercritical helium which is compressed above the critical pressure of 228 kPa has a stable cooling effect since the thermal conductivity does not change due to the evaporation of liquid helium. In order to maintain the temperature of the supercritical helium below 4 K before it enters a magnet, a heat exchanger is used. The compressor that IHI has developed has the ability to reduce the vapor pressure of liquid helium from atmospheric pressure to 50.7 kPa, and can attain the temperature of 3.5 K. The specification of this single stage centrifugal compressor is: mass flow rate 25 - 64 g/s, speed 80,000 rpm, adiabatic efficiency 62 - 69 %. The structure and the performance are reported. (K.I.)

  1. Helium refrigerator for 'SULTAN'

    International Nuclear Information System (INIS)

    Arpagaus, M.; Erlach, H.; Quack, H.

    1984-01-01

    The authors describe the helium refrigerator designed for the SULTAN test facility. SULTAN (Supraleiter-Testanlage) is intended to serve for the developments and testing of high field superconducting magnets. These magnets are needed mainly for future applications in nuclear fusion. (Auth.)

  2. Cosmological helium production simplified

    International Nuclear Information System (INIS)

    Bernstein, J.; Brown, L.S.; Feinberg, G.

    1988-01-01

    We present a simplified model of helium synthesis in the early universe. The purpose of the model is to explain clearly the physical ideas relevant to the cosmological helium synthesis, in a manner that does not overlay these ideas with complex computer calculations. The model closely follows the standard calculation, except that it neglects the small effect of Fermi-Dirac statistics for the leptons. We also neglect the temperature difference between photons and neutrinos during the period in which neutrons and protons interconvert. These approximations allow us to express the neutron-proton conversion rates in a closed form, which agrees to 10% accuracy or better with the exact rates. Using these analytic expressions for the rates, we reduce the calculation of the neutron-proton ratio as a function of temperature to a simple numerical integral. We also estimate the effect of neutron decay on the helium abundance. Our result for this quantity agrees well with precise computer calculations. We use our semi-analytic formulas to determine how the predicted helium abundance varies with such parameters as the neutron life-time, the baryon to photon ratio, the number of neutrino species, and a possible electron-neutrino chemical potential. 19 refs., 1 fig., 1 tab

  3. An overview of helium engineering demonstration loop (HENDEL) 1982

    International Nuclear Information System (INIS)

    Izawa, Naoki; Inagaki, Yoshiyuki; Tanaka, Toshiyuki; Hishida, Makoto; Shimomura, Hiroaki; Okamoto, Yoshizo

    1982-09-01

    The Helium Engineering Demonstration Loop (HENDEL) for proof testing of a large scale model of the VHTR components under simulated reactor operation conditions has been constructed at JAERI. This paper presented is markedly focused on the detail description of HENDEL facilities, including Mother (M), Adapter (A) and Test sections (T 1 -- T 4 ). (author)

  4. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  5. Helium production technology based on natural gas combustion and beneficial use of thermal energy

    Directory of Open Access Journals (Sweden)

    Nakoryakov Vladimir E.

    2016-01-01

    Full Text Available Helium is widely used in all industries, including power plant engineering. In recent years, helium is used in plants operating by the Brayton cycle, for example, in the nuclear industry. Using helium-xenon mixture in nuclear reactors has a number of advantages, and this area is rapidly developing. The hydrodynamics and mass transfer processes in single tubes with various cross-sections as well as in inter-channel space of heating tube bundle were studied at the Institute of Thermophysics, Siberian Branch of the Russian Academy of Sciences. Currently, there is a strongest shortage in helium production. The main helium production method consists in the liquefaction of the natural gas and subsequent separation of helium from remaining gas with its further purification using membranes.

  6. Antiprotonic helium atomcules

    Directory of Open Access Journals (Sweden)

    Sauge Sébastien

    2012-10-01

    Full Text Available About 3% of antiprotons ( stopped in helium are long-lived with microsecond lifetimes, against picoseconds in all other materials. This unusual longevity has been ascribed to the trapping of on metastable bound states in He+ helium atom-molecules thus named atomcules. Apart from their unique dual structure investigated by laser spectroscopy – a near-circular quasi-classical Rydberg atom with l ~ n – 1 ~ 37 or a special diatomic molecule with a negatively charged nucleus in high rotational state with J = l – the chemical physics aspects of their interaction with other atoms or molecules constitute an interesting topic for molecular physics. While atomcules may resist to million collisions in helium, molecular contaminants such as H2 are likely to destroy them in a single one, down to very low temperatures. In the Born-Oppenheimer framework, we interpret the molecular interaction obtained by ab initio quantum chemical calculations in terms of classical reactive channels, with activation barriers accounting for the experiments carried out in He and H2. From classical trajectory Monte Carlo simulations, we show that the thermalization stage strongly quenches initial populations, thus reduced to a recovered 3 % trapping fraction. This work illustrates the pertinence of chemical physics concepts to the study of exotic processes involving antimatter. New insights into the physico-chemistry of cold interstellar radicals are anticipated.

  7. High temperature embrittlement of metals by helium

    International Nuclear Information System (INIS)

    Schroeder, H.

    1983-01-01

    The present knowledge of the influence of helium on the high temperature mechanical properties of metals to be used as structural materials in fast fission and in future fusion reactors is reviewed. A wealth of experimental data has been obtained by many different experimental techniques, on many different alloys, and on different properties. This review is mostly concentrated on the behaviour of austenitic alloys -especially austenitic stainless steels, for which the data base is by far the largest - and gives only a few examples of special bcc alloys. The effect of the helium embrittlement on the different properties - tensile, fatigue and, with special emphasis, creep - is demonstrated by representative results. A comparison between data obtained from in-pile (-beam) experiments and from post-irradiation (-implantation) experiments, respectively, is presented. Theoretical models to describe the observed phenomena are briefly outlined and some suggestions are made for future work to resolve uncertainties and differences between our experimental knowledge and theoretical understanding of high temperature helium embrittlement. (author)

  8. Helium refrigeration system for hydrogen liquefaction applications

    Science.gov (United States)

    Nair, J. Kumar, Sr.; Menon, RS; Goyal, M.; Ansari, NA; Chakravarty, A.; Joemon, V.

    2017-02-01

    Liquid hydrogen around 20 K is used as cold moderator for generating “cold neutron beam” in nuclear research reactors. A cryogenic helium refrigeration system is the core upon which such hydrogen liquefaction applications are built. A thermodynamic process based on reversed Brayton cycle with two stage expansion using high speed cryogenic turboexpanders (TEX) along with a pair of compact high effectiveness process heat exchangers (HX), is well suited for such applications. An existing helium refrigeration system, which had earlier demonstrated a refrigeration capacity of 470 W at around 20 K, is modified based on past operational experiences and newer application requirements. Modifications include addition of a new heat exchanger to simulate cryogenic process load and two other heat exchangers for controlling the temperatures of helium streams leading out to the application system. To incorporate these changes, cryogenic piping inside the cold box is suitably modified. This paper presents process simulation, sizing of new heat exchangers as well as fabrication aspects of the modified cryogenic process piping.

  9. Interdiffusion of krypton and xenon in high-pressure helium

    International Nuclear Information System (INIS)

    Campana, R.J.; Jensen, D.D.; Epstein, B.D.; Hudson, R.G.; Baldwin, N.L.

    1980-01-01

    The interdiffusion of gaseous fission products in high-pressure helium is an important factor in the control of radioactivity in gas-cooled fast breeder reactors (GCFRs). As presently conceived, GCFRs use pressure-equalized and vented fuel in which fission gases released from the solid matrix oxide fuel are transported through the fuel rod interstices and internal fission product traps to the fuel assembly vents, where they are swept away to external traps and storage. Since the predominant transport process under steady-state operating conditions is interdiffusion of gaseous fission products in helium, the diffusion properties of krypton-helium and xenon-helium couples have been measured over the range of GCFR temperature and pressure conditions ( -1 ) and expected temperature dependence to the 1.66 power (Tsup(1.66)) at lower pressures and temperatures. Additional work is in progress to measure the behaviour of the krypton-helium and xenon-helium couples in GCFR fuel rod charcoal delay traps. (author)

  10. Helium-induced weld degradation of HT-9 steel

    International Nuclear Information System (INIS)

    Wang, Chin-An; Chin, B.A.; Lin, Hua T.; Grossbeck, M.L.

    1992-01-01

    Helium-bearing Sandvik HT-9 ferritic steel was tested for weldability to simulate the welding of structural components of a fusion reactor after irradiation. Helium was introduced into HT-9 steel to 0.3 and 1 atomic parts per million (appm) by tritium doping and decay. Autogenous single pass full penetration welds were produced using the gas tungsten arc (GTA) welding process under laterally constrained conditions. Macroscopic examination showed no sign of any weld defect in HT-9 steel containing 0.3 appm helium. However, intergranular micro cracks were observed in the HAZ of HT-9 steel containing 1 appm helium. The microcracking was attributed to helium bubble growth at grain boundaries under the influence of high stresses and temperatures that were present during welding. Mechanical test results showed that both yield strength (YS) and ultimate tensile strength (UTS) decreased with increasing temperature, while the total elongation increased with increasing temperature for all control and helium-bearing HT-9 steels

  11. HEINBE; the calculation program for helium production in beryllium under neutron irradiation

    International Nuclear Information System (INIS)

    Shimakawa, Satoshi; Ishitsuka, Etsuo; Sato, Minoru

    1992-11-01

    HEINBE is a program on personal computer for calculating helium production in beryllium under neutron irradiation. The program can also calculate the tritium production in beryllium. Considering many nuclear reactions and their multi-step reactions, helium and tritium productions in beryllium materials irradiated at fusion reactor or fission reactor may be calculated with high accuracy. The calculation method, user's manual, calculated examples and comparison with experimental data were described. This report also describes a neutronics simulation method to generate additional data on swelling of beryllium, 3,000-15,000 appm helium range, for end-of-life of the proposed design for fusion blanket of the ITER. The calculation results indicate that helium production for beryllium sample doped lithium by 50 days irradiation in the fission reactor, such as the JMTR, could be achieved to 2,000-8,000 appm. (author)

  12. Preliminary Overview of a Helium Cooling System for the Secondary Helium Loop in VHTR-based SI Hydrogen Production Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Cho, Mintaek; Kim, Dahee; Lee, Taehoon; Lee, Kiyoung; Kim, Yongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear hydrogen production facilities consist of a very high temperature gas-cooled nuclear reactor (VHTR) system, intermediate heat exchanger (IHX) system, and a sulfur-iodine (SI) thermochemical process. This study focuses on the coupling system between the IHX system and SI thermochemical process. To prevent the propagation of the thermal disturbance owing to the abnormal operation of the SI process components from the IHX system to the VHTR system, a helium cooling system for the secondary helium of the IHX is required. In this paper, the helium cooling system has been studied. The temperature fluctuation of the secondary helium owing to the abnormal operation of the SI process was then calculated based on the proposed coupling system model. Finally, the preliminary conceptual design of the helium cooling system with a steam generator and forced-draft air-cooled heat exchanger to mitigate the thermal disturbance has been carried out. A conceptual flow diagram of a helium cooling system between the IHX and SI thermochemical processes in VHTR-based SI hydrogen production facilities has been proposed. A helium cooling system for the secondary helium of the IHX in this flow diagram prevents the propagation of the thermal disturbance from the IHX system to the VHTR system, owing to the abnormal operation of the SI process components. As a result of a dynamic simulation to anticipate the fluctuations of the secondary helium temperature owing to the abnormal operation of the SI process components with a hydrogen production rate of 60 mol·H{sub 2}/s, it is recommended that the maximum helium cooling capacity to recover the normal operation temperature of 450 .deg. C is 31,933.4 kJ/s. To satisfy this helium cooling capacity, a U-type steam generator, which has a heat transfer area of 12 m{sup 2}, and a forced-draft air-cooled condenser, which has a heat transfer area of 12,388.67 m{sup 2}, are required for the secondary helium cooling system.

  13. Surface electrons of helium films

    International Nuclear Information System (INIS)

    Studart, N.; Hipolito, O.

    1986-01-01

    Theoretical calculations of some properties of two-dimensional electrons on a liquid helium film adsorbed on a solid substrate are reviewed. We describe the spectrum of electron bound states on bulk helium as well on helium films. The correlational properties, such as the structure factor and correlation energy, are determined as functions of the film thickness for different types of substrates in the framework of a Generalized Random-Phase Approximation. The collective excitations of this system are also described. The results for electrons on the surface of thin films and bulk helium are easily obtained. we examine the electron interaction with the excitations of the liquid helium surface resulting in a new polaron state, which was observed very recently. The ground state energy and the effective mass of this polaron are determined by using the path-integral formalism and unitary-transformation method. Recent speculations about the phase diagram of electrons on the helium film are also discussed. (Author) [pt

  14. Status of helium-cooled nuclear power systems. [Development potential

    Energy Technology Data Exchange (ETDEWEB)

    Melese-d' Hospital, G.; Simnad, M

    1977-09-01

    Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor is also discussed. The status of helium-cooled nuclear energy systems is summarized as a basis for assessing their prospects. 50 references.

  15. Process for detecting leak faults using a helium mass spectrometer

    International Nuclear Information System (INIS)

    Divet, Claude; Morin, Claude.

    1977-01-01

    The description is given of a process for detecting very small leak faults putting into communication the outer and inner sides of the wall of a containment, one of these wall sides being in contact with gaseous helium under a pressure of around one torr, the other side being one of the limits of a space pumped down to a residual gas pressure under 10 -3 torr. This space is in communication with the measuring cell of a helium mass spectrometer. This process may be applied to the detection of faults in metal claddings of the fuel rods used in nuclear reactors [fr

  16. Canada's helium output rising fast

    Energy Technology Data Exchange (ETDEWEB)

    1966-12-01

    About 12 months from now, International Helium Limited will be almost ready to start up Canada's second helium extraction plant at Mankota, in Saskatchewan's Wood Mountain area about 100 miles southwest of Moose Jaw. Another 80 miles north is Saskatchewan's (and Canada's) first helium plant, operated by Canadian Helium and sitting on a gas deposit at Wilhelm, 9 miles north of Swift Current. It contains almost 2% helium, some COD2U, and the rest nitrogen. One year in production was apparently enough to convince Canadian Helium that the export market (it sells most of its helium in W. Europe) can take a lot more than it's getting. Construction began this summer on an addition to the Swift Current plant that will raise its capacity from 12 to 36MMcf per yr when it goes on stream next spring. Six months later, International Helium's 40 MMcf per yr plant to be located about 4 miles from its 2 Wood Mountain wells will double Canada's helium output again.

  17. Orion A helium abundance

    International Nuclear Information System (INIS)

    Tsivilev, A.P.; Ershov, A.A.; Smirnov, G.T.; Sorochenko, R.L.

    1986-01-01

    The 22.4-GHz (H,He)66-alpha and 36.5-GHz (H,He)56-alpha radio recombination lines have been observed at several Jaffe-Pankonin positions in the central part of the Orion A source. The measured relative abundance of ionized helium increases with distance, averaging 11.6 percent at peripheral points. The observed behavior is interpreted by a blister-type model nebula, which implies that Orion A has a true He abundance of 12 percent, is moving with a radial velocity of 5 km/sec, and is expanding. 18 references

  18. Simulation of liquid helium

    International Nuclear Information System (INIS)

    Ceperley, D.M.

    1985-07-01

    The author discusses simulation methods for quantum mechanical systems at finite temperatures. Recently it has been shown that static properties of some quantum systems can be obtained by simulation in a straightforward manner using path integrals, albeit with an order of magnitude more computing effort needed than for the corresponding classical systems. Some dynamical information can be gleaned from these simulations as will be discussed below. But this is very limited - there is no quantum version of the molecular dynamics method. The path integral method is illustrated by discussing the application to liquid helium. 12 refs., 8 figs

  19. High Efficiency Regenerative Helium Compressor, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Helium plays several critical rolls in spacecraft propulsion. High pressure helium is commonly used to pressurize propellant fuel tanks. Helium cryocoolers can be...

  20. A helium regenerative compressor

    International Nuclear Information System (INIS)

    Swift, W.L.; Nutt, W.E.; Sixsmith, H.

    1994-01-01

    This paper discusses the design and performance of a regenerative compressor that was developed primarily for use in cryogenic helium systems. The objectives for the development were to achieve acceptable efficiency in the machine using conventional motor and bearing technology while reducing the complexity of the system required to control contamination from the lubricants. A single stage compressor was built and tested. The compressor incorporates aerodynamically shaped blades on a 218 mm (8.6 inches) diameter impeller to achieve high efficiency. A gas-buffered non-contact shaft seal is used to oppose the diffusion of lubricant from the motor bearings into the cryogenic circuit. Since it is a rotating machine, the flow is continuous and steady, and the machine is very quiet. During performance testing with helium, the single stage machine has demonstrated a pressure ratio of 1.5 at a flow rate of 12 g/s with measured isothermal efficiencies in excess of 30%. This performance compares favorably with efficiencies generally achieved in oil flooded screw compressors

  1. Installation and Commissioning of the Helium Refrigeration System for the HANARO-CNS

    International Nuclear Information System (INIS)

    Choi, Jung Woon; Kim, Young Ki; Wu, Sang Ik; Son, Woo Jung

    2009-11-01

    The cold neutron source (CNS), which will be installed in the vertical CN hole of the reflector tank at HANARO, makes thermal neutrons to moderate into the cold neutrons with the ranges of 0.1 ∼ 10 meV passing through a moderator at about 22K. A moderator to produce cold neutrons is liquid hydrogen, which liquefies by the heat transfer with cryogenic helium flowing from the helium refrigeration system. For the maintenance of liquid hydrogen in the IPA, the CNS system is mainly consisted of the hydrogen system to supply the hydrogen to the IPA, the vacuum system to keep the cryogenic liquid hydrogen in the IPA, and the helium refrigeration system to liquefy the hydrogen gas. The helium refrigeration system can be divided into two sections: one is the helium compression part from the low pressure gas to the high pressure gas and the other is the helium expansion part from the high temperature gas and pressure to low temperature and pressure gas by the expansion turbine. The helium refrigeration system except the warm helium pipe and the helium buffer tank has been manufactured by Linde Kryotechnik, AG in Switzerland and installed in the research reactor hall, HANARO. Other components have been manufactured in the domestic company. This technical report deals with the issues, its solutions, and other particular points while the helium refrigeration system was installed at site, verified its performance, and conducted its commissioning along the reactor operation. Furthermore, the operation procedure of the helium refrigeration system is included in here for the normal operation of the CNS

  2. Helium production measurements for neutron dosimetry and damage correlations

    International Nuclear Information System (INIS)

    Farrar, H. IV; Lippincott, E.P.

    1978-01-01

    Helium accumulation fluence monitors (HAFM's), consisting of miniature vanadium capsules containing small, accurately-known amounts of 10 B or 6 Li, are being used routinely for neutron dosimetry measurements in breeder reactor environments. Additionally, solid wires of Al, Fe and Cu have been irradiated by 14.8-MeV neutrons from the d-T reaction, and measurements of the helium production along these wires have given detailed neutron fluence profiles. Additional materials with relatively high (n,α) cross sections are being tested in a wide variety of neutron environments to select HAFM sets that will provide spectral information by unfolding techniques. The mass spectrometric helium measurement technique has been demonstrated to produce results with better than 2% (1 sigma) absolute accuracy. Intercomparisons with other laboratories have demonstrated good correlations with radiometric and fission chamber dosimetry results

  3. Helium leak testing of large pressure vessels or subassemblies

    International Nuclear Information System (INIS)

    Hopkins, J.S.; Valania, J.J.

    1977-01-01

    Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

  4. Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

    Science.gov (United States)

    Mansur, L. K.; Grossbeck, M. L.

    1988-07-01

    Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

  5. Ab initio Investigation of Helium in Vanadium Oxide Nanoclusters

    Science.gov (United States)

    Danielson, Thomas; Tea, Eric; Hin, Celine

    Nanostructured ferritic alloys (NFAs) are strong candidate materials for the next generation of fission reactors and future fusion reactors. They are characterized by a large number density of oxide nanoclusters dispersed throughout a BCC iron matrix, where current oxide nanoclusters are primarily comprised of Y-Ti-O compounds. The oxide nanoclusters provide the alloy with high resistance to neutron irradiation, high yield strength and high creep strength at the elevated temperatures of a reactor environment. In addition, the oxide nanoclusters serve as trapping sites for transmutation product helium providing substantially increased resistance to catastrophic cracking and embrittlement. Although the mechanical properties and radiation resistance of the existing NFAs is promising, the problem of forming large scale reactor components continues to present a formidable challenge due to the high hardness and unpredictable fracture behavior of the alloys. An alternative alloy has been previously proposed and fabricated where vanadium is added in order to form vanadium oxide nanoclusters that serve as deflection sites for crack propagation. Although experiments have shown evidence that the fracture behavior of the alloys is improved, it is unknown whether or not the vanadium oxide nanoclusters are effective trapping sites for helium. We present results obtained using density functional theory investigating the thermodynamic stability of helium with the vanadium oxide matrix to make a comparison of trapping effectiveness to traditional Y-Ti-O compounds.

  6. Helium transport and exhaust studies in enhanced confinement regimes in DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Burrell, K.H.; Finkenthal, D.F.; Gohil, P.; Groebner, R.J.

    1995-02-01

    A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust (τ* He /τ E ∼ 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D He /X eff ∼1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system

  7. New technique for enhancing helium production in ferritic materials

    International Nuclear Information System (INIS)

    Greenwood, L.R.; Graczyk, D.G.; Kneff, D.W.

    1987-10-01

    Analyses of iron samples irradiated up to 10 27 n/m 2 in HFIR found more helium than was expected from fast neutron reactions at high neutron fluences. The helium excess increases systematically with neutron exposure, suggesting a transmutation-driven process. The extra helium may be produced in two different ways, either by fast neutron reactions on the transmuted isotopes of iron or by a thermal neutron reaction with the radioactive isotope 55 Fe. Radiometric and mass spectrometric measurements of the iron isotopes composing the irradiated samples have been used to determine limits on the cross sections for each process. Either of these processes can be used to enhance helium production in ferritic materials during irradiations in mixed-spectrum reactors by isotopically enriching the samples. Further work is needed to clarify the reaction mechanisms and helium production cross sections. Our measurements determined the thermal neutron total absorption cross section of 55 Fe to be 13.2 +- 2.1 barns. 16 refs., 3 figs., 3 tabs

  8. Photoionization of helium dimers

    International Nuclear Information System (INIS)

    Havermeier, Tilo

    2010-01-01

    The helium dimer is one of the most weakly bound systems in the universe. This makes it an interesting quantum mechanical object for investigation. These Van der Waals Clusters can be produced in an expansion of a cryogenic gas jet through a small nozzle into vacuum. In the present experiment we examine the interaction of He dimers with synchrotron radiation at an energy range from 64 to 78 eV. We observed different pathways leading to single ionization of both He atoms of the dimer compound. This two close standing ions begin now to dissociate in cause of their coulomb potential. All charged fragments were detected in coincidence with a COLTRIMS system. Especially Interatomic Coulombic Decay (ICD) and the two step process (TS1) were clearly identified. Furthermore a distribution of the internuclear distance was obtained from the measured Kinetic Energy Release (KER). (orig.)

  9. Helium generation and diffusion in graphite and some carbides

    International Nuclear Information System (INIS)

    Holt, J.B.; Guinan, M.W.; Hosmer, D.W.; Condit, R.H.; Borg, R.J.

    1976-01-01

    The cross section for the generation of helium in neutron irradiated carbon was found to be 654 mb at 14.4 MeV and 744 mb at 14.9 MeV. Extrapolating to 14.1 MeV (the fusion reactor spectrum) gives 615 mb. The diffusion of helium in dense polycrystalline graphite and in pyrographite was measured and found to be D = 7.2 x 10 -7 m 2 s -1 exp (-80 kJ/RT). It is assumed that diffusion is primarily in the basal plane direction in crystals of the graphite. In polycrystalline graphite the path length is a factor of √2 longer than the measured distance due to the random orientation mismatch between successive grains. Isochronal anneals (measured helium release as the specimen is steadily heated) were run and maximum release rates were found at 200 0 C in polycrystalline graphite, 1000 0 C in pyrographite, 1350 0 C in boron carbide, and 1350 0 and 2400 0 C (two peaks) in silicon carbide. It is concluded that in these candidates for curtain materials in fusion reactors the helium releases can probably occur without bubble formation in graphites, may occur in boron carbide, but will probably cause bubble formation in silicon carbide. 7 figures

  10. Exotic helium molecules

    International Nuclear Information System (INIS)

    Portier, M.

    2007-12-01

    We study the photo-association of an ultracold cloud of magnetically trapped helium atoms: pairs of colliding atoms interact with one or two laser fields to produce a purely long range 4 He 2 (2 3 S 1 -2 3 P 0 ) molecule, or a 4 He 2 (2 3 S 1 -2 3 S 1 ) long range molecule. Light shifts in one photon photo-association spectra are measured and studied as a function of the laser polarization and intensity, and the vibrational state of the excited molecule. They result from the light-induced coupling between the excited molecule, and bound and scattering states of the interaction between two metastable atoms. Their analysis leads to the determination of the scattering length a = (7.2 ± 0.6) ruling collisions between spin polarized atoms. The two photon photo-association spectra show evidence of the production of polarized, long-range 4 He 2 (2 3 S 1 -2 3 S 1 ) molecules. They are said to be exotic as they are made of two metastable atoms, each one carrying a enough energy to ionize the other. The corresponding lineshapes are calculated and decomposed in sums and products of Breit-Wigner and Fano profiles associated to one and two photon processes. The experimental spectra are fit, and an intrinsic lifetime τ = (1.4 ± 0.3) μs is deduced. It is checked whether this lifetime could be limited by spin-dipole induced Penning autoionization. This interpretation requires that there is a quasi-bound state close to the dissociation threshold in the singlet interaction potential between metastable helium atoms for the theory to match the experiment. (author)

  11. Helium turbomachinery operating experience from gas turbine power plants and test facilities

    International Nuclear Information System (INIS)

    McDonald, Colin F.

    2012-01-01

    The closed-cycle gas turbine, pioneered and deployed in Europe, is not well known in the USA. Since nuclear power plant studies currently being conducted in several countries involve the coupling of a high temperature gas-cooled nuclear reactor with a helium closed-cycle gas turbine power conversion system, the experience gained from operated helium turbomachinery is the focus of this paper. A study done as early as 1945 foresaw the use of a helium closed-cycle gas turbine coupled with a high temperature gas-cooled nuclear reactor, and some two decades later this was investigated but not implemented because of lack of technology readiness. However, the first practical use of helium as a gas turbine working fluid was recognized for cryogenic processes, and the first two small fossil-fired helium gas turbines to operate were in the USA for air liquefaction and nitrogen production facilities. In the 1970's a larger helium gas turbine plant and helium test facilities were built and operated in Germany to establish technology bases for a projected future high efficiency large nuclear gas turbine power plant concept. This review paper covers the experience gained, and the lessons learned from the operation of helium gas turbine plants and related test facilities, and puts these into perspective since over three decades have passed since they were deployed. An understanding of the many unexpected events encountered, and how the problems, some of them serious, were resolved is important to avoid them being replicated in future helium turbomachines. The valuable lessons learned in the past, in many cases the hard way, particularly from the operation in Germany of the Oberhausen II 50 MWe helium gas turbine plant, and the technical know-how gained from the formidable HHV helium turbine test facility, are viewed as being germane in the context of current helium turbomachine design work being done for future high efficiency nuclear gas turbine plant concepts. - Highlights:

  12. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Parish, C.M., E-mail: parishcm@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Unocic, K.A.; Tan, L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zinkle, S.J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Kondo, S. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto, 611-0011 (Japan); Snead, L.L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Hoelzer, D.T.; Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2017-01-15

    We irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ∼50 dpa, ∼15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ∼8 nm, ∼10{sup 21} m{sup −3} (CNA), and of ∼3 nm, 10{sup 23} m{sup −3} (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ∼50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.

  13. Corrosion behaviour of high temperature alloys in impure helium environments

    International Nuclear Information System (INIS)

    Shindo, Masami; Quadakkers, W.J.; Schuster, H.

    1986-01-01

    Corrosion tests with Ni-base high temperature alloys were carried out at 900 and 950 0 C in simulated high temperature reactor helium environments. It is shown that the carburization and decarburization behaviour is strongly affected by the Cr and Ti(Al) contents of the alloys. In carburizing environments, additions of Ti, alone or in combination with Al, significantly improve the carburization resistance. In oxidizing environment, the alloys with high Cr and Al(Ti) contents are the most resistant against decarburization. In this environment alloys with additions of Ti and Al show poor oxidation resistance. The experimental results obtained are compared with a recently developed theory describing corrosion of high temperature alloys in high temperature reactor helium environments. (orig.)

  14. Design study on the helium engineering demonstration loop (HENDEL)

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Muto, Yasushi; Suzuki, Kunihiko

    1977-11-01

    Four reference studies made on Helium Engineering Demonstration Loop (HENDEL) are described. HENDEL is used in confirmation of the designs of VHTR components such as reactor structure, core structure, intermediate heat exchanger and piping. It consists of mother loop, adapter section and four test sections for fuel stack, reactor support and insulation structure, core structure and high temperature heat transfer component respectively. System and component designs of the mother and adapter section and preliminary designs of the four test sections are shown. And, the plans of operation, instrumentation, control, safety, utilities (electricity, cooling water and helium gas) and construction schedule of HENDEL and research and development of the test sections are also briefed. (auth.)

  15. Electronic properties of physisorbed helium

    International Nuclear Information System (INIS)

    Kossler, Sarah

    2011-01-01

    This thesis deals with electronic excitations of helium physisorbed on metal substrates. It is studied to what extent the electronic properties change compared to the gas phase due to the increased helium density and the proximity of the metal. Furthermore, the influence of different substrate materials is investigated systematically. To this end, up to two helium layers were adsorbed onto Ru (001), Pt (111), Cu (111), and Ag (111) surfaces in a custom-made cryostat. These samples were studied spectroscopically using synchrotron radiation and a time-of-flight detector. The experimental results were then analyzed in comparison with extensive theoretical model calculations.

  16. Electronic properties of physisorbed helium

    Energy Technology Data Exchange (ETDEWEB)

    Kossler, Sarah

    2011-09-22

    This thesis deals with electronic excitations of helium physisorbed on metal substrates. It is studied to what extent the electronic properties change compared to the gas phase due to the increased helium density and the proximity of the metal. Furthermore, the influence of different substrate materials is investigated systematically. To this end, up to two helium layers were adsorbed onto Ru (001), Pt (111), Cu (111), and Ag (111) surfaces in a custom-made cryostat. These samples were studied spectroscopically using synchrotron radiation and a time-of-flight detector. The experimental results were then analyzed in comparison with extensive theoretical model calculations.

  17. Baking method for thermonuclear reactor

    International Nuclear Information System (INIS)

    Kobayashi, Shigetada.

    1986-01-01

    Purpose: To improve the heat transmission property to the reactor core structures thereby shortening the baking time for the reactor core in thermonuclear reactors. Constitution: High temperature airs are supplied from a baking system to cooling pipeways disposed within reactor core structures and helium gas is supplied from a helium gas supply system through the reactor core structures to the inside of the reactor core for scavenging. The scavenging operation may be combined with vacuum suction. Further, the inside of the reactor is scavenged while maintaining at such a negative pressure as within a range not degrading the heat conduction property. Since the helium gas is chemically inert and poor in the depositing property, it shows no adsorbability even for the material heated to high temperature. Further, since the diffusion and heat conduction properties are high, the heat conduction property to the materials upon baking can be improved to shorten the baking time. No disadvantages are caused by the introduction of the helium gas upon baking. (Kawakami, Y.)

  18. Summary report on technical experiences from high-temperature helium turbomachinery testing in Germany

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1996-01-01

    In Germany a comprehensive research and development program was initiated in 1968 for a Brayton (closed) cycle power conversion system. The program was for ultimate use with a high temperature, helium cooled nuclear reactor heat source (the HHT project) for electricity generation using helium as the working fluid. The program continued until 1982 in international cooperation with the United States and Switzerland. This document describes the designs and reports the results of testing activities that addressed the development of turbines, compressors, hot gas ducts, materials, heat exchangers and other equipment items for use with a helium working fluid at high temperatures. 67 refs, 34 figs, tabs

  19. A reassessment of the effects of helium on Charpy impact properties of ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Gelles, D.S.; Hamilton, M.L.; Hankin, G.L.

    1998-01-01

    To test the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of tests performed on specimens irradiated in reactors with very different neutron spectra, and isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400 C to 100 dpa and 1000 appm He will result in a ductile to brittle transition temperature shift of over 500 C. However, the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations also support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in HFIR. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation

  20. Characterization of high flux magnetized helium plasma in SCU-PSI linear device

    Science.gov (United States)

    Xiaochun, MA; Xiaogang, CAO; Lei, HAN; Zhiyan, ZHANG; Jianjun, WEI; Fujun, GOU

    2018-02-01

    A high-flux linear plasma device in Sichuan University plasma-surface interaction (SCU-PSI) based on a cascaded arc source has been established to simulate the interactions between helium and hydrogen plasma with the plasma-facing components in fusion reactors. In this paper, the helium plasma has been characterized by a double-pin Langmuir probe. The results show that the stable helium plasma beam with a diameter of 26 mm was constrained very well at a magnetic field strength of 0.3 T. The core density and ion flux of helium plasma have a strong dependence on the applied current, magnetic field strength and gas flow rate. It could reach an electron density of 1.2 × 1019 m-3 and helium ion flux of 3.2 × 1022 m-2 s-1, with a gas flow rate of 4 standard liter per minute, magnetic field strength of 0.2 T and input power of 11 kW. With the addition of -80 V applied to the target to increase the helium ion energy and the exposure time of 2 h, the flat top temperature reached about 530 °C. The different sizes of nanostructured fuzz on irradiated tungsten and molybdenum samples surfaces under the bombardment of helium ions were observed by scanning electron microscopy. These results measured in the SCU-PSI linear device provide a reference for International Thermonuclear Experimental Reactor related PSI research.

  1. Operating Manual of Helium Refrigerator (Rev. 2)

    Energy Technology Data Exchange (ETDEWEB)

    Song, K.M.; Son, S.H.; Kim, K.S.; Lee, S.K.; Kim, M.S. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    A helium refrigerator was installed as a supplier of 20K cold helium to the cryogenic distillation system of WTRF pilot plant. The operating procedures of the helium refrigerator, helium compressor and auxiliary apparatus are described for the safety and efficient operation in this manual. The function of the helium refrigerator is to remove the impurities from the compressed helium of about 250psig, to cool down the helium from ambient temperature to 20K through the heat exchanger and expansion engine and to transfer the cold helium to the cryogenic distillation system. For the smoothly operation of helium refrigerator, the preparation, the start-up, the cool-down and the shut-down of the helium refrigerator are described in this operating manual. (author). 3 refs., 14 tabs.

  2. Tribological behavior of zirconium coatings in high temperature helium

    International Nuclear Information System (INIS)

    Cachon, Lionel; Albaladejo, Serge; Taraud, Pascal

    2005-01-01

    In France, a comprehensive research and development program is leaded by the CEA, since 2001, for the Gas Cooled Reactor (GCR) project using helium as cooling fluid, in order to establish the feasibility of the technology of an early VHTR prototype to be started by 2015, and then to qualify the generic VHTR technology, so as to meet similar objectives for the GFR. In this frame a tribology program has been launched. The purpose of the work presented in this paper is to describe the CEA Helium tribology study: high temperature gas cooled reactors require wear protection (thermal barriers, control rod drive mechanisms, reactor internals, ...). Tests in helium atmosphere are necessary to be fully representative of tribological environments and finally to check the possible materials or coatings which can provide a reliable answer to these situations. The main characteristics and first experimental results are thus described. This paper focus on tribology tests leaded in the temperature range 800-1000degC, on ceramic (ZrO 2 -Y 2 O 3 ) with and without solid lubricant like CaF2). (author)

  3. Laser spectroscopy of antiprotonic helium

    CERN Document Server

    Hori, M

    2005-01-01

    When antiprotons (i.e. the antimatter counterpart of protons) are stopped in helium gas, 97% of them annihilate within picoseconds by reacting with the helium nuclei; a 3% fraction, however, survive with an anomalously long lifetime of several microseconds. This longevity is due to the formation of antiprotonic helium, which is a three-body Rydberg atom composed of an antiproton, electron, and helium nucleus. The ASACUSA experimental collaboration has recently synthesized large numbers of these atoms using CERN's Antiproton Decelerator facility, and measured the atom's transition frequencies to 60 parts per billion by laser spectroscopy. By comparing the experimental results with recent three-body QED calculations and the known antiproton cyclotron frequency, we were able to show that the antiproton mass and charge are the same as the corresponding proton values to a precision of 10 parts per billion. Ongoing and future series of experiments will further improve the experimental precision by using chirp-compe...

  4. High Accuracy Vector Helium Magnetometer

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed HAVHM instrument is a laser-pumped helium magnetometer with both triaxial vector and omnidirectional scalar measurement capabilities in a single...

  5. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  6. Helium diffusion in irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1981-03-01

    Boron carbide has been internationally adopted as the neutron absorber material in the control and safety rods of large fast breeder reactors. Its relatively large neutron capture cross section at high neutron energies provides sufficient reactivity worth with a minimum of core space. In addition, the commercial availability of boron carbide makes it attractive from a fabrication standpoint. Instrumented irradiation experiments in EBR-II have provided continuous helium release data on boron carbide at a variety of operating temperatures. Although some microstructural and compositional variations were examined in these experiments most of the boron carbide was prototypic of that used in the Fast Flux Test Facility. The density of the boron carbide pellets was approximately 92% of theoretical. The boron carbide pellets were approximately 1.0 cm in diameter and possessed average grain sizes that varied from 8 to 30 μm. Pellet centerline temperatures were continually measured during the irradiation experiments

  7. High temperature tensile properties of 316 stainless steel implanted with helium

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Yamamoto, Norikazu; Shiraishi, Haruki

    1993-01-01

    Helium embrittlement is one of the problems in structural materials for fusion reactors. Recently, martensitic steels have been developed which have a good resistance to high-temperature helium embrittlement, but the mechanism has not yet been clarified. In this paper, tensile behaviors of helium implanted austenitic stainless steels, which are sensitive to the helium embrittlement, were studied and compared with those of martensitic steels under the same experimental conditions, and the effect of microstructure on helium embrittlement was discussed. Helium was implanted by 300 appm at 573-623 K to miniature tensile speciments of 316 austenitic steels using a cyclotron accelerator. Solution annealed (316SA) and 20% cold worked (316CW) specimens were used. Post-implantation tensile tests were carried out at 573, 873 and 973 K. Yield stress at 573 K increased with the helium implantation in 316SA and 316CW, but the yield stress changes of 316SA at 873 and 973 K were different from that of 316CW. Black-dots were observed in the as-implanted specimen and bubbles were observed in the speciments tensile-tested at 873 and 973 K. Intergranular fracture was observed at only 973 K in both of the 316SA and 316CW specimens. Therefore, cold work did not suppress the high-temperature helium embrittlement under this experimental condition. The difference in the influence of helium on type 316 steel and 9Cr martensitic steels were discussed. Test temperature change of reduction in are showed clearly that helium embrittlement did not occur in 9Cr martensitic steels but occurred in 316 austenitic steels. Fine microstructures of 9Cr martensitic steels should suppress helium embrittlement at high temperatures. (author)

  8. Some research and development on power plants with helium gas turbine units

    International Nuclear Information System (INIS)

    Kaplan, M.P.

    1983-01-01

    Research and development projects pursued at the S. M. Kirov Kharkov Turbine Factory Production Association for Nuclear Turbomachinery Manufacturing have probed into prospective use of helium as the working medium in nuclear power facilities. The projects under study are compared mainly in terms of heat efficiency. Solutions are also being sought for problems centering around high efficiency in helium turbocompressors combined with shortening of axial dimensions. Different types of power plants are being compared with attention given to features of the flow passages of turbocompressors. The projects were developed for helium temperatures and pressures downstream of the reactor 950 0 C and 4.8 MPa, and thermal reactor output 2250 MW(th). The reactor is assumed to be served by two turbine plants in the turbocompressor designs

  9. Cryogenic filter method produces super-pure helium and helium isotopes

    Science.gov (United States)

    Hildebrandt, A. F.

    1964-01-01

    Helium is purified when cooled in a low pressure environment until it becomes superfluid. The liquid helium is then filtered through iron oxide particles. Heating, cooling and filtering processes continue until the purified liquid helium is heated to a gas.

  10. Microstructural evolution in dual-ion irradiated 316SS under various helium injection schedules

    International Nuclear Information System (INIS)

    Kohyama, A.; Igata, N.; Ayrault, G.; Tokyo Univ.

    1984-01-01

    Dual-ion irradiated 316 SS samples with various helium injection schedules were studied. The intent of using different schedules was to either approximate the MFR condition, mimic the mixed spectrum reactor condition or mimic the fast reactor condition. The objective of this investigation is to study the influence of these different helium injection schedules on the microstructural development under irradiation. The materials for this study was 316 SS (MFE heat) with three thermomechanical pre-irradiation treatments: solution annealed, solution annealed and aged and 20% cold worked. The cavity nucleation and growth stages were investigated using high resolution TEM. (orig.)

  11. Maintenance free gas bearing helium blower for nuclear plant

    Science.gov (United States)

    Molyneaux, A., Dr; Harris, M., Prof; Sharkh, S., Prof; Hill, S.; de Graaff, T.

    2017-08-01

    This paper describes the design, testing and operation of novel helium blowers used to recirculate the helium blanketing gas in the nuclear reactor used as a neutron source at the Institut Laue Langevan, Grenoble, France. The laser sintered shrouded centrifugal wheel operates at speeds up to 45000 rpm supported on helium lubricated hydrodynamic spiral groove bearings, and is driven by a sensorless permanent magnet motor. The entire machine is designed to keep the helium gas (polluted by a small amount of D2O) out of contact with any iron or copper materials which would contribute to the corrosion of parts of the circuit. It is designed to have zero maintenance during a lifetime of 40,000 hours of continuous operation. This paper will describe the spiral groove journal and thrust bearings. Design and manufacture of the 1 kW motor and centrifugal wheel will be explained including their CFD and FEA analyses. Measurements of rotor displacement will be presented showing the behaviour under factory testing as well as details of the measured centrifugal wheel and motor performances. Two machines are incorporated into the circuit to provide redundancy and the first blower has been in continuous operation since Jan 2015. The blower was designed, manufactured, assembled and tested in the UK using predominantly UK suppliers.

  12. Helium bubbles aggravated defects production in self-irradiated copper

    Science.gov (United States)

    Wu, FengChao; Zhu, YinBo; Wu, Qiang; Li, XinZhu; Wang, Pei; Wu, HengAn

    2017-12-01

    Under the environment of high radiation, materials used in fission and fusion reactors will internally accumulate numerous lattice defects and bubbles. With extensive studies focused on bubble resolution under irradiation, the mutually effects between helium bubbles and displacement cascades in irradiated materials remain unaddressed. Therefore, the defects production and microstructure evolution under self-irradiation events in vicinity of helium bubbles are investigated by preforming large scale molecular dynamics simulations in single-crystal copper. When subjected to displacement cascades, distinguished bubble resolution categories dependent on bubble size are observed. With the existence of bubbles, radiation damage is aggravated with the increasing bubble size, represented as the promotion of point defects and dislocations. The atomic mechanisms of heterogeneous dislocation structures are attributed to different helium-vacancy cluster modes, transforming from the resolved gas trapped with vacancies to the biased absorption of vacancies by the over-pressured bubble. In both cases, helium impedes the recombination of point defects, leading to the accelerated formation of interstitial loops. The results and insight obtained here might contribute to understand the underlying mechanism of transmutant solute on the long-term evolution of irradiated materials.

  13. IBA studies of helium mobility in nuclear materials revisited

    Energy Technology Data Exchange (ETDEWEB)

    Trocellier, P., E-mail: patrick.trocellier@cea.fr [CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France); Agarwal, S.; Miro, S. [CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France); Vaubaillon, S. [CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France); CEA, INSTN, UEPTN, F-91191 Gif-sur-Yvette (France); Leprêtre, F.; Serruys, Y. [CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France)

    2015-12-15

    The aim of this paper is to point out and to discuss some features extracted from the study of helium migration in nuclear materials performed during the last fifteen years using ion beam analysis (IBA) measurements. The first part of this paper is devoted to a brief description of the two main IBA methods used, i.e. deuteron induced nuclear reaction for {sup 3}He depth profiling and high-energy heavy-ion induced elastic recoil detection analysis for {sup 4}He measurement. In the second part, we provide an overview of the different studies carried out on model nuclear waste matrices and model nuclear reactor structure materials in order to illustrate and discuss specific results in terms of key influence parameters in relation with thermal or radiation activated migration of helium. Finally, we show that among the key parameters we have investigated as able to influence the height of the helium migration barrier, the following can be considered as pertinent: the experimental conditions used to introduce helium (implanted ion energy and implantation fluence), the grain size of the matrix, the lattice cell volume, the Young's modulus, the ionicity degree of the chemical bond between the transition metal atom M and the non-metal atom X, and the width of the band gap.

  14. Impact of neutron irradiation on thermal helium desorption from iron

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

    2017-06-15

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  15. Numerical simulation of losses along a natural circulation helium loop

    Energy Technology Data Exchange (ETDEWEB)

    Knížat, Branislav, E-mail: branislav.knizat@stuba.sk; Urban, František, E-mail: frantisek.urban@stuba.sk; Mlkvik, Marek, E-mail: marek.mlkvik@stuba.sk; Ridzoň, František, E-mail: frantisek.ridzon@stuba.sk; Olšiak, Róbert, E-mail: robert.olsiak@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Mechanical Engineering, Nám. slobody 17, 812 31 Bratislava, Slovak Republik (Slovakia)

    2016-06-30

    A natural circulation helium loop appears to be a perspective passive method of a nuclear reactor cooling. When designing this device, it is important to analyze the mechanism of an internal flow. The flow of helium in the loop is set in motion due to a difference of hydrostatic pressures between cold and hot branch. Steady flow at a requested flow rate occurs when the buoyancy force is adjusted to resistances against the flow. Considering the fact that the buoyancy force is proportional to a difference of temperatures in both branches, it is important to estimate the losses correctly in the process of design. The paper deals with the calculation of losses in branches of the natural circulation helium loop by methods of CFD. The results of calculations are an important basis for the hydraulic design of both exchangers (heater and cooler). The analysis was carried out for the existing model of a helium loop of the height 10 m and nominal heat power 250 kW.

  16. Apparatus to measure low level helium for neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ozaki, Shuji; Takao, Yoshiyuki; Muramasu, Masatomo; Hida, Tomoya; Sou, Hirofumi; Nakashima, Hideki [Kyushu Univ., Fukuoka (Japan); Kanda, Yukinori

    1998-03-01

    An apparatus to measure low level helium in a solid sample for neutron dosimetry in the practical use such as area monitoring in the long-term and reactor surveillance was reported. In our previous work, the helium atoms measurement system (HAMS) was developed. A sample was evaporated in the furnace and the released gas from the sample was analyzed with the mass spectrometer of the system to determine the amount of helium contained in it. The system has been improved to advance the lower helium measurement limit in a solid sample for its application to an area monitoring system. The mass of a solid is up to 100mg. Two important points should be considered to advance the lower limit. One was to produce a high quality vacuum in the system chamber for suppressing background gases during the sample measurement. The other important point was to detect very small output from the mass spectrometer. A pulse counting system was used to get high sensitivity in the mass 4 analyzing. (author)

  17. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.

    1989-07-01

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  18. Current status and future development of modular high temperature gas cooled reactor technology

    International Nuclear Information System (INIS)

    2001-02-01

    This report includes an examination of the international activities with regard to the development of the modular HTGR coupled to a gas turbine. The most significant of these gas turbine programmes include the pebble bed modular reactor (PBMR) being designed by ESKOM of South Africa and British Nuclear Fuels plc. (BNFL) of the United Kingdom, and the gas turbine-modular helium reactor (GT-MHR) by a consortium of General Atomics of the United States of America, MINATOM of the Russian Federation, Framatome of France and Fuji Electric of Japan. Details of the design, economics and plans for these plants are provided in Chapters 3 and 4, respectively. Test reactors to evaluate the safety and general performance of the HTGR and to support research and development activities including electricity generation via the gas turbine and validation of high temperature process heat applications are being commissioned in Japan and China. Construction of the high temperature engineering test reactor (HTTR) by the Japan Atomic Energy Research Institute (JAERI) at its Oarai Research Establishment has been completed with the plant currently in the low power physics testing phase of commissioning. Construction of the high temperature reactor (HTR-10) by the Institute of Nuclear Energy Technology (INET) in Beijing, China, is nearly complete with initial criticality expected in 2000. Chapter 5 provides a discussion of purpose, status and testing programmes for these two plants. In addition to the activities related to the above mentioned plants, Member States of the IWGGCR continue to support research associated with HTGR safety and performance as well as development of alternative designs for commercial applications. These activities are being addressed by national energy institutes and, in some projects, private industry, within China, France, Germany, Indonesia, Japan, the Netherlands, the Russian Federation, South Africa, United Kingdom and the USA. Chapter 6 includes details

  19. Elevated temperature and high pressure large helium gas loop

    International Nuclear Information System (INIS)

    Sakasai, Minoru; Midoriyama, Shigeru; Miyata, Toyohiko; Nakase, Tsuyoshi; Izaki, Makoto

    1979-01-01

    The development of high temperature gas-cooled reactors especially aiming at the multi-purpose utilization of nuclear heat energy is carried out actively in Japan and West Germany. In Japan, the experimental HTGR of 50 MWt and 1000 deg C outlet temperature is being developed by Japan Atomic Energy Research Institute and others since 1969, and the development of direct iron-making technology utilizing high temperature reducing gas was started in 1973 as the large project of Ministry of Internalional Trade and Industry. Kawasaki Heavy Industries, Ltd., Has taken part in these development projects, and has developed many softwares for nuclear heat design, system design and safety design of nuclear reactor system and heat utilization system. In hardwares also, efforts have been exerted to develop the technologies of design and manufacture of high temperature machinery and equipments. The high temperature, high pressure, large helium gas loop is under construction in the technical research institute of the company, and it is expected to be completed in December, 1979. The tests planned are that of proving the dynamic performances of the loop and its machinery and equipments and the verification of analysis codes. The loop is composed of the main circulation system, the objects of testing, the helium gas purifying system, the helium supplying and evacuating system, instruments and others. (Kako, I.)

  20. Pierre Gorce working on a helium pump.

    CERN Multimedia

    1975-01-01

    This type of pump was designed by Mario Morpurgo, to circulate liquid helium in superconducting magnets wound with hollow conductors. M. Morpurgo, Design and construction of a pump for liquid helium, CRYIOGENICS, February 1977, p. 91

  1. Convective mixing in helium white dwarfs

    International Nuclear Information System (INIS)

    Vauclair, G.; Fontaine, G.

    1979-01-01

    The conditions under which convective mixing episodes take place between the helium envelopes and the underlying carbon layers in helium-rich white dwarfs are investigated. It is found that, for essentially any value of the initial helium content less than the maximum mass a helium convection zone can have, mixing does occur, and leads, in the vast majority of cases, to an almost pure carbon superficial composition. Mixing products that show only traces of carbon while retaining helium-dominated envelopes are possible only if the initial helium content is quite close to the maximum possible mass of the helium convection zone. In the presence of turbulence, this restriction could be relaxed, however, and the helium-rich lambda4670 stars may possibly be explained in this fashion

  2. The consequences of helium production on microstructural development in isotopically tailored ferritic alloys

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1996-01-01

    A series of alloys have been made adding various isotopes of nickel in order to vary the production of helium during irradiation by a two step nuclear reaction in a mixed spectrum reactor. The alloys use a base composition of Fe-12Cr with an addition of 1.5% nickel, either in the form of 60 Ni which produces no helium, 59 Ni which produces helium at a rate of about 10 appm He/dpa, or natural nickel ( Nat Ni) which provides an intermediate level of helium due to delayed development of 59 Ni. Specimens were irradiated in the HFIR at Oak Ridge, TN to ∼7 dpa at 300 and 400 degrees C. Microstructural examinations indicated that nickel additions promote precipitation in all alloys, but the effect appears to be much stronger at 400 degrees C than at 300 degrees C. There is sufficient dose by 7 dpa (and with 2 appm He) to initiate void swelling in ferritic/martensitic alloys. Little difference was found between response from 59 Ni and Nat Ni. Also, helium bubble development for high helium generation conditions appeared to be very different at 300 and 400 degrees C. At 300 degrees C, it appeared that high densities of bubbles formed whereas at 400 degrees C, bubbles could not be identified, possibly because of the complexity of the microstructure, but more likely because helium accumulated at precipitate interfaces

  3. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hee Joon [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Edwards, Dan J., E-mail: dan.edwards@pnnl.gov [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Kurtz, Richard J. [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Yamamoto, Takuya; Wu, Yuan [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Odette, G. Robert [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Materials Department, University of California, Santa Barbara, CA 93106 (United States)

    2017-02-15

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron {sup 58}Ni(n{sub th},γ) {sup 59}Ni(n{sub th},α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al{sub 2}YO{sub 3} oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  4. The consequences of helium production on microstructural development in isotopically tailored ferritic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest Lab., Richland, WA (United States)

    1996-10-01

    A series of alloys have been made adding various isotopes of nickel in order to vary the production of helium during irradiation by a two step nuclear reaction in a mixed spectrum reactor. The alloys use a base composition of Fe-12Cr with an addition of 1.5% nickel, either in the form of {sup 60}Ni which produces no helium, {sup 59}Ni which produces helium at a rate of about 10 appm He/dpa, or natural nickel ({sup Nat}Ni) which provides an intermediate level of helium due to delayed development of {sup 59}Ni. Specimens were irradiated in the HFIR at Oak Ridge, TN to {approx}7 dpa at 300 and 400{degrees}C. Microstructural examinations indicated that nickel additions promote precipitation in all alloys, but the effect appears to be much stronger at 400{degrees}C than at 300{degrees}C. There is sufficient dose by 7 dpa (and with 2 appm He) to initiate void swelling in ferritic/martensitic alloys. Little difference was found between response from {sup 59}Ni and {sup Nat}Ni. Also, helium bubble development for high helium generation conditions appeared to be very different at 300 and 400{degrees}C. At 300{degrees}C, it appeared that high densities of bubbles formed whereas at 400{degrees}C, bubbles could not be identified, possibly because of the complexity of the microstructure, but more likely because helium accumulated at precipitate interfaces.

  5. Helium localization around the microscopic impurities embedded to liquid helium

    International Nuclear Information System (INIS)

    Gordon, E.B.; Shestakov, A.F.

    2000-01-01

    The structure and properties of the environment round the impurity atoms (Im) embedded in liquid helium are considered. It is shown that there are two qualitatively different types of structure of the He atom layer next to Im - attraction and repulsion structures. For the center attraction structure (strong Im-He interaction) the Im-He separation is longer than the equilibrium one for the pair Im-He potential, and the density and localization of He atoms are higher than in the bulk. It this case the He atom content in the layer, n, is almost independent of applied pressure. In the repulsion structure realized for alkaline metal atoms the Im-He separation is shorter than the equilibrium one and the density is lower than in the helium bulk. At T approx 1 K occupied are several states with different n and their energies differ only by approx 0.1 K, an increase in pressure resulting in a considerable reduction of n. The optical and EPR spectra of the atoms embedded to liquid and solid helium are interpreted on the basis of the analysis carried out. A simple model is proposed to evaluate the helium surroundings characteristics from the experimental pressure dependences of atomic line shifts in the absorption and emission spectra. The attraction structures in 3 He - 4 He mixtures are suggested to be highly enriched by 4 He atoms which the repulsion structures - by 3 He atoms. a possibility for existence of phase transitions in helium shells surrounding impurity atoms is considered

  6. Wide-range vortex shedding flowmeter for high-temperature helium gas

    Energy Technology Data Exchange (ETDEWEB)

    Baker, S.P.; Herndon, P.G.; Ennis, R.M. Jr.

    1983-01-01

    The existing design of a commercially available vortex shedding flowmeter (VSFM) was modified and optimized to produce three 4-in. and one 6-in. high-performance VSFMs for measuring helium flow in a gas-cooled fast reactor (GCFR) test loop. The project was undertaken because of the significant economic and performance advantages to be realized by using a single flowmeter capable of covering the 166:1 flow range (at 350/sup 0/C and 45:1 pressure range) of the tests. A detailed calibration in air and helium at the Colorado Engineering Experiment Station showed an accuracy of +-1% of reading for a 100:1 helium flow range and +-1.75% of reading for a 288:1 flow range in both helium and air. At an extended gas temperature of 450/sup 0/C, water cooling was necessary for reliable flowmeter operation.

  7. Hybrid simulation research on formation mechanism of tungsten nanostructure induced by helium plasma irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Atsushi M., E-mail: ito.atsushi@nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Takayama, Arimichi; Oda, Yasuhiro [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Tamura, Tomoyuki; Kobayashi, Ryo; Hattori, Tatsunori; Ogata, Shuji [Nagoya Institute of Technology, Gokiso-cho, Showa-ku, Nagoya 466-8555 (Japan); Ohno, Noriyasu; Kajita, Shin [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Yajima, Miyuki [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Noiri, Yasuyuki [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Yoshimoto, Yoshihide [University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Saito, Seiki [Kushiro National College of Technology, Kushiro, Hokkaido 084-0916 (Japan); Takamura, Shuichi [Aichi Institute of Technology, 1247 Yachigusa, Yakusa-cho, Toyota 470-0392 (Japan); Murashima, Takahiro [Tohoku University, 6-3, Aramaki-Aza-Aoba, Aoba-Ward, Sendai 980-8578 (Japan); Miyamoto, Mitsutaka [Shimane University, Matsue, Shimane 690-8504 (Japan); Nakamura, Hiroaki [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

    2015-08-15

    The generation of tungsten fuzzy nanostructure by exposure to helium plasma is one of the important problems for the use of tungsten material as divertor plates in nuclear fusion reactors. In the present paper, the formation mechanisms of the helium bubble and the tungsten fuzzy nanostructure were investigated by using several simulation methods. We proposed the four-step process which is composed of penetration step, diffusion and agglomeration step, helium bubble growth step, and fuzzy nanostructure formation step. As the fourth step, the formation of the tungsten fuzzy nanostructure was successfully reproduced by newly developed hybrid simulation combining between molecular dynamics and Monte-Carlo method. The formation mechanism of tungsten fuzzy nanostructure observed by the hybrid simulation is that concavity and convexity of the surface are enhanced by the bursting of helium bubbles in the region around the concavity.

  8. High Temperature Operational Experiences of Helium Experimental Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung-Deok; Kim, Eung-Seon; Kim, Min Hwan

    2015-01-01

    The development of high temperature components of VHTR is very important because of its higher operation temperature than that of a common light water reactor and high pressure industrial process. The development of high temperature components requires the large helium loop. Many countries have high temperature helium loops or a plan for its construction. Table 1 shows various international state-of-the-art of high temperature and high pressure gas loops. HELP performance test results show that there is no problem in operation of HELP at the very high temperature experimental condition. These experimental results also provide the basic information for very high temperature operation with bench-scale intermediate heat exchanger prototype in HELP. In the future, various heat exchanger tests will give us the experimental data for GAMMA+ validation about transient T/H behavior of the IHX prototype and the optimization of the working fluid in the intermediate loop

  9. Defects induced by helium implantation in SiC

    International Nuclear Information System (INIS)

    Oliviero, E.; Barbot, J.F.; Declemy, A.; Beaufort, M.F.; Oliviero, E.

    2008-01-01

    SiC is one of the considered materials for nuclear fuel conditioning and for the fabrication of some core structures in future nuclear generation reactors. For the development of this advance technology, a fundamental research on this material is of prime importance. In particular, the implantation/irradiation effects have to be understood and controlled. It is with this aim that the structural alterations induced by implantation/irradiation in SiC are studied by different experimental techniques as transmission electron microscopy, helium desorption, X-ray diffraction and Rutherford backscattering spectrometry. In this work, the different types of defects induced by helium implantation in SiC, point or primary defects (obtained at low energy (∼100 eV) until spread defects (obtained at higher energy (until ∼2 MeV)) are exposed. The amorphization/recrystallization and swelling phenomena are presented too. (O.M.)

  10. Explosive helium burning in white dwarf stars

    Energy Technology Data Exchange (ETDEWEB)

    Khokhlov, A.M. (AN SSSR, Moscow. Astronomicheskij Sovet)

    1984-04-01

    Helium burning kinetics in white dwarfs has been considered at constant temperatures T >= 10/sup 9/ K and densities rho >10/sup 5/ g/cm/sup 3/. It is found, that helium detonation in white dwarfs does not lead to formation of light (A < 56) elements. Thus, helium white dwarf model for supernova 1 is inconsistent with observations.

  11. Clustering of Helium Atoms at a ½

    NARCIS (Netherlands)

    Berg, F. v.d.; Heugten, W. v.; Caspers, L.M.; Veen, A. v.; Hosson, J.Th.M. de

    1977-01-01

    Atomistic calculations on a ½<111>{110} edge dislocation show a restricted tendency of clustering of helium atom along this dislocation. Clusters with up to 4 helium atoms have been studied. A cluster with 3 helium proved to be most stable.

  12. Muonium and neutral muonic helium

    International Nuclear Information System (INIS)

    Orth, H.

    1981-01-01

    In this brief article the current status on muonium spectroscopy with emphasis on recent developments will be summarized. The experimental and theoretical progress of the muonic helium atom will be reviewed. Future directions in this field of research will be discussed. (orig./HSI)

  13. Electric response in superfluid helium

    Czech Academy of Sciences Publication Activity Database

    Chagovets, Tymofiy

    2016-01-01

    Roč. 488, May (2016), s. 62-66 ISSN 0921-4526 R&D Projects: GA ČR GP13-03806P Institutional support: RVO:68378271 Keywords : superfluid helium * electric response * second sound * ions in He II Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.386, year: 2016

  14. Production of negative helium ions

    International Nuclear Information System (INIS)

    Toledo, A.S. de; Sala, O.

    1977-01-01

    A negative helium ion source using potassium charge exchange vapor has been developed to be used as an injector for the Pelletron accelerator. 3 He and α beam currents of up to 2μA have been extracted with 75% particle transmission through the machine [pt

  15. Crystal orientation effects on helium ion depth distributions and adatom formation processes in plasma-facing tungsten

    International Nuclear Information System (INIS)

    Hammond, Karl D.; Wirth, Brian D.

    2014-01-01

    We present atomistic simulations that show the effect of surface orientation on helium depth distributions and surface feature formation as a result of low-energy helium plasma exposure. We find a pronounced effect of surface orientation on the initial depth of implanted helium ions, as well as a difference in reflection and helium retention across different surface orientations. Our results indicate that single helium interstitials are sufficient to induce the formation of adatom/substitutional helium pairs under certain highly corrugated tungsten surfaces, such as (1 1 1)-orientations, leading to the formation of a relatively concentrated layer of immobile helium immediately below the surface. The energies involved for helium-induced adatom formation on (1 1 1) and (2 1 1) surfaces are exoergic for even a single adatom very close to the surface, while (0 0 1) and (0 1 1) surfaces require two or even three helium atoms in a cluster before a substitutional helium cluster and adatom will form with reasonable probability. This phenomenon results in much higher initial helium retention during helium plasma exposure to (1 1 1) and (2 1 1) tungsten surfaces than is observed for (0 0 1) or (0 1 1) surfaces and is much higher than can be attributed to differences in the initial depth distributions alone. The layer thus formed may serve as nucleation sites for further bubble formation and growth or as a source of material embrittlement or fatigue, which may have implications for the formation of tungsten “fuzz” in plasma-facing divertors for magnetic-confinement nuclear fusion reactors and/or the lifetime of such divertors.

  16. Helium generation reaction rates for 6Li and 10B in benchmark facilities

    International Nuclear Information System (INIS)

    Farrar, Harry IV; Oliver, B.M.; Lippincott, E.P.

    1980-01-01

    The helium generation rates for 10 B and 6 Li have been measured in two benchmark reactor facilities having neutron spectra similar to those found in a breeder reactor. The irradiations took place in the Coupled Fast Reactivity Measurements Facility (CFRMF) and in the 10% enriched 235 U critical assembly, BIG-10. The helium reaction rates were obtained by precise high-sensitivity gas mass spectrometric analyses of the helium content of numerous small samples. Comparison of these reaction rates with other reaction rates measured in the same facilities, and with rates calculated from published cross sections and from best estimates of the neutron spectral shapes, indicate significant discrepancies in the calculated values. Additional irradiations in other benchmark facilities have been undertaken to better determine the energy ranges where the discrepancies lie

  17. Simplicity works for superfluid helium

    International Nuclear Information System (INIS)

    Bowley, Roger

    2000-01-01

    The famous philosopher Karl Popper once said that ''science is the art of systematic oversimplification''. Indeed, when faced with a new puzzle the trick is to simplify it without losing the essential physics - something that is easier said than done. However, this approach has paid off recently in low-temperature physics. Last year Richard Packard, Seamus Davis and co-workers at the University of California at Berkeley encountered a puzzling new phenomenon in superfluid helium-3, a quantum fluid that remains a liquid close to absolute zero and exhibits unusual properties such as the ability to flow without friction (A Machenkov et al. 1999 Phys. Rev. Lett. 83 3860). Previous experiments had revealed that certain effects in liquid helium are analogous to effects observed in superconductors, materials that lose all resistance to electric current at low temperatures. When the Berkeley researchers connected two reservoirs of superfluid helium-3, the superfluid flowed back and forth through apertures that formed a ''weak link'' between the two containers. This behaviour is similar to the oscillatory current of electrons that can flow across an insulating gap separating two superconductors - a device that is known as a Josephson junction. What was puzzling about the Berkeley results was that the helium-3 had two different stable configurations, both of which behaved in an unconventional way compared with a Josephson junction. This puzzle has now been solved independently by Sidney Yip at the National Center for Theoretical Sciences in Taiwan, and by Janne Viljas and Erkki Thuneberg at the Helsinki University of Technology in Finland (Phys. Rev. Lett. 1999 83 3864 and 3868). In this article the author describes the latest research on superfluid helium. (UK)

  18. Simplicity works for superfluid helium

    Energy Technology Data Exchange (ETDEWEB)

    Bowley, Roger [University of Nottingham, Nottingham (United Kingdom)

    2000-02-01

    The famous philosopher Karl Popper once said that ''science is the art of systematic oversimplification''. Indeed, when faced with a new puzzle the trick is to simplify it without losing the essential physics - something that is easier said than done. However, this approach has paid off recently in low-temperature physics. Last year Richard Packard, Seamus Davis and co-workers at the University of California at Berkeley encountered a puzzling new phenomenon in superfluid helium-3, a quantum fluid that remains a liquid close to absolute zero and exhibits unusual properties such as the ability to flow without friction (A Machenkov et al. 1999 Phys. Rev. Lett. 83 3860). Previous experiments had revealed that certain effects in liquid helium are analogous to effects observed in superconductors, materials that lose all resistance to electric current at low temperatures. When the Berkeley researchers connected two reservoirs of superfluid helium-3, the superfluid flowed back and forth through apertures that formed a ''weak link'' between the two containers. This behaviour is similar to the oscillatory current of electrons that can flow across an insulating gap separating two superconductors - a device that is known as a Josephson junction. What was puzzling about the Berkeley results was that the helium-3 had two different stable configurations, both of which behaved in an unconventional way compared with a Josephson junction. This puzzle has now been solved independently by Sidney Yip at the National Center for Theoretical Sciences in Taiwan, and by Janne Viljas and Erkki Thuneberg at the Helsinki University of Technology in Finland (Phys. Rev. Lett. 1999 83 3864 and 3868). In this article the author describes the latest research on superfluid helium. (UK)

  19. Effect of HTGR helium on fatigue and creep properties of 2 1/4Cr-1Mo steel

    International Nuclear Information System (INIS)

    Kurumaji, T.; Yamazaki, H.; Kudo, A.

    1982-01-01

    Low cycle fatigue and creep tests have been carried out on 2 1/4Cr-1Mo steel (candidate steel for VHTR reactor pressure vessel) in helium environment containing 200 approx. 300 μatm of H 2 , 100 approx. 150 μatm CO, 7 approx. 10 μatm CH 4 , 7 approx. 10 μatm CO 2 and 1 μatm H 2 O (JAERI B Helium). Fatigue life in helium environment was longer than that in air at 450 0 C. This results can be explained by supposing that oxidation at the crack tip causes the wedge effect to promote crack propagation in air. On the otherhand, creep rupture strength showed no significant difference in both helium and air. Equivalent creep rupture strength in both helium and air may be due to the fact that detrimental internal oxidation and carburization or decarburization hardly occur at 400 approx. 450 0 C

  20. 'In-beam' simulation of high temperature helium embrittlement of DIN 1.4970 austenitic stainless steel

    International Nuclear Information System (INIS)

    Schroeder, H.; Batfalsky, P.

    1982-01-01

    This work describes a facility for high temperature creep rupture tests during homogeneous helium implantation. This 'in-beam' creep testing facility is used to simulate helium embrittlement effects which will be very important for first wall materials of future fusion reactors operated at high temperatures. First results for DIN 1.4970 austenitic stainless steel clearly demonstrate differences between samples 'in-beam' tested at 1073 K and those creep tested at the same temperature after room temperature helium implantation. The specimens ruptured 'in-beam' have much shorter lifetimes and lower ductility than the specimens tested after room temperature implantation. There are also differences in the microstructures, concerning helium bubble sizes and densities in matrix and grain boundaries. These microstructural differences may be a key for the understanding of the more severe helium embrittlement effects 'in-beam' as compared to creep tests performed after room temperature implantation. (orig.)

  1. Fusion reactor high vacuum pumping

    International Nuclear Information System (INIS)

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M.

    1992-01-01

    This paper reports on recent experiments which have shown the practicality of using activated carbon (coconut charcoal) at 4K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were satisfactory. The long-term effects of tritium on the charcoal/cement system developed by Grumman and LLNL was now known; therefore a program was undertaken to see what, if any, effect long-term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77 K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately one-third way through, and after the exposure. Modest effects were noted which would not seriously restrict the use of charcoal as a cryosorber for fusion reactor high-vacuum pumping applications

  2. Hydrogen and helium trapping in tungsten deposition layers formed by RF plasma sputtering

    International Nuclear Information System (INIS)

    Kazunari Katayama; Kazumi Imaoka; Takayuki Okamura; Masabumi Nishikawa

    2006-01-01

    Understanding of tritium behavior in plasma facing materials is an important issue for fusion reactor from viewpoints of fuel control and radiation safety. Tungsten is used as a plasma facing material in the divertor region of ITER. However, investigation of hydrogen isotope behavior in tungsten deposition layer is not sufficient so far. It is also necessary to evaluate an effect of helium on a formation of deposition layer and an accumulation of hydrogen isotopes because helium generated by fusion reaction exists in fusion plasma. In this study, tungsten deposition layers were formed by sputtering method using hydrogen and helium RF plasma. An erosion rate and a deposition rate of tungsten were estimated by weight measurement. Hydrogen and helium retention were investigated by thermal desorption method. Tungsten deposition was performed using a capacitively-coupled RF plasma device equipped with parallel-plate electrodes. A tungsten target was mounted on one electrode which is supplied with RF power at 200 W. Tungsten substrates were mounted on the other electrode which is at ground potential. The plasma discharge was continued for 120 hours where pressure of hydrogen or helium was controlled to be 10 Pa. The amounts of hydrogen and helium released from deposition layers was quantified by a gas chromatograph. The erosion rate of target tungsten under helium plasma was estimated to be 1.8 times larger than that under hydrogen plasma. The deposition rate on tungsten substrate under helium plasma was estimated to be 4.1 times larger than that under hydrogen plasma. Atomic ratio of hydrogen to tungsten in a deposition layer formed by hydrogen plasma was estimated to be 0.17 by heating to 600 o C. From a deposition layer formed by helium plasma, not only helium but also hydrogen was released by heating to 500 o C. Atomic ratios of helium and hydrogen to tungsten were estimated to be 0.080 and 0.075, respectively. The trapped hydrogen is probably impurity hydrogen

  3. Testing of degradation of alloy 800 H in impure helium at 760 °C

    Czech Academy of Sciences Publication Activity Database

    Berka, J.; Vilémová, Monika; Sajdl, P.

    2015-01-01

    Roč. 464, September (2015), s. 221-229 ISSN 0022-3115 Institutional support: RVO:61389021 Keywords : High temperature corrosion * impure helium * 800 H * Generation IV nuclear reactors Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311515002019#

  4. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Moons, F.

    1998-01-01

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  5. HeREF-2003: Helium Refrigeration Techniques

    CERN Multimedia

    2003-01-01

    CERN Technical Training 2003: Learning for the LHC ! Theory, Technology, Maintenance and Control of Helium Refrigerators HeREF-2003 is a course in the framework of the 2002 Technical Training Programme, that will provide a complete introduction to Helium refrigeration, with a practical approach to theory, technology, maintenance and control of Helium refrigeration installations. Theoretical aspects and equations will be limited to a minimum. HeREF-2003 targets an audience of technicians and operators of Helium refrigeration plants at CERN, as well as physicists and engineers needing an overview of current Helium refrigeration techniques. HeREF-2003 will address, among other, issues related to component technology, installation maintenance, process control and Helium purity. A commented visit to a couple of CERN Helium refrigeration or liquefaction plants will also take place. Duration: 7 half days (4 mornings and 3 afternoons), 6-10 October, 2003 Cost per participant: 500.- CHF Language: Bilingual English...

  6. HeREF-2003 : Helium Refrigeration Techniques

    CERN Multimedia

    2003-01-01

    CERN Technical Training 2003: Learning for the LHC ! Theory, Technology, Maintenance and Control of Helium Refrigerators HeREF-2003 is a course in the framework of the 2002 Technical Training Programme, that will provide a complete introduction to Helium refrigeration, with a practical approach to theory, technology, maintenance and control of Helium refrigeration installations. Theoretical aspects and equations will be limited to a minimum. HeREF-2003 targets an audience of technicians and operators of Helium refrigeration plants at CERN, as well as physicists and engineers needing an overview of current Helium refrigeration techniques. HeREF-2003 will address, among other, issues related to component technology, installation maintenance, process control and Helium purity. A commented visit to a couple of CERN Helium refrigeration or liquefaction plants will also take place. • Duration: 7 half days (4 mornings and 3 afternoons), 6-10 October, 2003 • Cost per participant: 500.- CHF ...

  7. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  8. MIT pebble bed reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

    2007-03-15

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

  9. MIT pebble bed reactor project

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2007-01-01

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

  10. Behavior of W-based materials in hot helium gas

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Vilémová, Monika; Hadraba, Hynek; Di Gabriele, F.; Kuběna, Ivo; Kolíbalová, E.; Michalička, J.; Čech, J.; Jäger, Aleš

    2016-01-01

    Roč. 9, December (2016), s. 405-410 ISSN 2352-1791. [International Conference of Fusion Reactor Material (ICFRM-17) /17./. Aachen, 11.10.2015-16.10.2015] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 ; RVO:68081723 ; RVO:68378271 Keywords : tungsten * helium * fusion materials Subject RIV: JG - Metallurgy; JG - Metallurgy (UFM-A); JG - Metallurgy (FZU-D) http://dx.doi.org/10.1016/j.nme.2016.03.009

  11. Prediction of TRISO coated particle performances for a one-pass deep burn

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-02-15

    In the present studies, TRISO coated particle performances have been investigated for incinerating plutonium and minor actinides by the Gas Turbine-Modular Helium Reactor, whose fresh fuel is fabricated after the uranium extraction (UREX) process applied to Light Water Reactors irradiated fuel. The analyses divide into two parts: in the first part, the latest design of the reactor core proposed by General Atomics, which takes advantage of four fuel rings, has been modeled in deep details by the Monte Carlo MCNP code and a burnup process has been simulated by the MCB code. In the second part, the TRISO coated particle performances have been investigated by the PANAMA code with the goal of verifying the design constraints proposed by General Atomics. During burnup, the refueling and shuffling schedule followed the one-pass deep burn concept, where the fuel is utilized, since fabrication for the Gas Turbine-Modular Helium Reactor, without any reprocessing until the final disposal into the geological repository. During the reactor operation, the fast fluence on all TRISO particles layers has been evaluated and the production of the key fission products monitored. During an hypothetical reactor accident scenario, the TRISO particle failure fraction has been estimated.

  12. Prediction of TRISO coated particle performances for a one-pass deep burn

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2008-01-01

    In the present studies, TRISO coated particle performances have been investigated for incinerating plutonium and minor actinides by the Gas Turbine-Modular Helium Reactor, whose fresh fuel is fabricated after the uranium extraction (UREX) process applied to Light Water Reactors irradiated fuel. The analyses divide into two parts: in the first part, the latest design of the reactor core proposed by General Atomics, which takes advantage of four fuel rings, has been modeled in deep details by the Monte Carlo MCNP code and a burnup process has been simulated by the MCB code. In the second part, the TRISO coated particle performances have been investigated by the PANAMA code with the goal of verifying the design constraints proposed by General Atomics. During burnup, the refueling and shuffling schedule followed the one-pass deep burn concept, where the fuel is utilized, since fabrication for the Gas Turbine-Modular Helium Reactor, without any reprocessing until the final disposal into the geological repository. During the reactor operation, the fast fluence on all TRISO particles layers has been evaluated and the production of the key fission products monitored. During an hypothetical reactor accident scenario, the TRISO particle failure fraction has been estimated

  13. Impulse approximation in solid helium

    International Nuclear Information System (INIS)

    Glyde, H.R.

    1985-01-01

    The incoherent dynamic form factor S/sub i/(Q, ω) is evaluated in solid helium for comparison with the impulse approximation (IA). The purpose is to determine the Q values for which the IA is valid for systems such a helium where the atoms interact via a potential having a steeply repulsive but not infinite hard core. For 3 He, S/sub i/(Q, ω) is evaluated from first principles, beginning with the pair potential. The density of states g(ω) is evaluated using the self-consistent phonon theory and S/sub i/(Q,ω) is expressed in terms of g(ω). For solid 4 He resonable models of g(ω) using observed input parameters are used to evaluate S/sub i/(Q,ω). In both cases S/sub i/(Q, ω) is found to approach the impulse approximation S/sub IA/(Q, ω) closely for wave vector transfers Q> or approx. =20 A -1 . The difference between S/sub i/ and S/sub IA/, which is due to final state interactions of the scattering atom with the remainder of the atoms in the solid, is also predominantly antisymmetric in (ω-ω/sub R/), where ω/sub R/ is the recoil frequency. This suggests that the symmetrization procedure proposed by Sears to eliminate final state contributions should work well in solid helium

  14. Liquid helium plant in Dubna

    International Nuclear Information System (INIS)

    Agapov, N.N.; Baldin, A.M.; Kovalenko, A.D.

    1995-01-01

    The liquid-helium cooling capacity installed at the Laboratory of High Energies is about 5 kw at a 4.5 K temperature level. It is provided with four industrial helium liquefiers of 1.6 kw/4.5 K each. They have been made by the Russian enterprise NPO GELYMASH and upgraded by the specialists of the Laboratory. The first one was put into operation in 1980, the two others in 1991, and the last one is under commissioning. The development of the LHE cryoplant was concerned with the construction of the new superconducting accelerator Nuclotron aimed to accelerate nuclei and heavy ions up to energies of 6 GeV/u. The first test run at the Nuclotron was carried out in March 1993, and the total running time has been about 2000 hours up to now. Since 1992 the cryoplant has been intensively used by the users outside the Laboratory. More than a million liters of liquid helium was provided in 1993 for such users. The reliability of the cryoplant system was as high as 98 percent for 4500 hours of operation in 1993-1994. 7 refs., 4 figs., 1 tab

  15. Cluster dynamics modeling of the effect of high dose irradiation and helium on the microstructure of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Brimbal, Daniel, E-mail: Daniel.brimbal@areva.com [AREVA NP, Tour AREVA, 1 Place Jean Millier, 92084 Paris La Défense (France); Fournier, Lionel [AREVA NP, Tour AREVA, 1 Place Jean Millier, 92084 Paris La Défense (France); Barbu, Alain [Alain Barbu Consultant, 6 Avenue Pasteur Martin Luther King, 78230 Le Pecq (France)

    2016-01-15

    A mean field cluster dynamics model has been developed in order to study the effect of high dose irradiation and helium on the microstructural evolution of metals. In this model, self-interstitial clusters, stacking-fault tetrahedra and helium-vacancy clusters are taken into account, in a configuration well adapted to austenitic stainless steels. For small helium-vacancy cluster sizes, the densities of each small cluster are calculated. However, for large sizes, only the mean number of helium atoms per cluster size is calculated. This aspect allows us to calculate the evolution of the microstructural features up to high irradiation doses in a few minutes. It is shown that the presence of stacking-fault tetrahedra notably reduces cavity sizes below 400 °C, but they have little influence on the microstructure above this temperature. The binding energies of vacancies to cavities are calculated using a new method essentially based on ab initio data. It is shown that helium has little effect on the cavity microstructure at 300 °C. However, at higher temperatures, even small helium production rates such as those typical of sodium-fast-reactors induce a notable increase in cavity density compared to an irradiation without helium. - Highlights: • Irradiation of steels with helium is studied through a new cluster dynamics model. • There is only a small effect of helium on cavity distributions in PWR conditions. • An increase in helium production causes an increase in cavity density over 500 °C. • The role of helium is to stabilize cavities via reduced emission of vacancies.

  16. Self-trapping of helium in metals

    International Nuclear Information System (INIS)

    Wilson, W.D.; Bisson, C.L.; Baskes, M.I.

    1981-01-01

    Atomistic calculations are presented which demonstrate that helium atoms in a metal lattice are able to cluster with each other, producing vacancies and nearby self-interstitial defects. Even a small number of helium atoms is found to be sufficient to create these large distortions. As few as five interstitial helium can spontaneously produce a lattice vacancy and nearby self-interstitial. An eight-helium-atom cluster gives rise to two such defects, and 16 helium atoms to more than five self-interstitial vacancy pairs. It was noted that the self-interstitials prefer to agglomerate on the same ''side'' of the helium cluster rather than to spread themselves out uniformly. The binding energy of each additional helium atom to these clusters increases with helium concentration and the trap is apparently unsaturable. A rate theory using these atomistic binding energies has been used to calculate the kinetics of helium-bubble nucleation and growth. The results are consistent with measurements of the properties of helium resulting from tritium decay

  17. Helium in inert matrix dispersion fuels

    International Nuclear Information System (INIS)

    Veen, A. van; Konings, R.J.M.; Fedorov, A.V.

    2003-01-01

    The behaviour of helium, an important decay product in the transmutation chains of actinides, in dispersion-type inert matrix fuels is discussed. A phenomenological description of its accumulation and release in CERCER and CERMET fuel is given. A summary of recent He-implantation studies with inert matrix metal oxides (ZrO 2 , MgAl 2 O 4 , MgO and Al 2 O 3 ) is presented. A general picture is that for high helium concentrations helium and vacancy defects form helium clusters which convert into over-pressurized bubbles. At elevated temperature helium is released from the bubbles. On some occasions thermal stable nano-cavities or nano-pores remain. On the basis of these results the consequences for helium induced swelling and helium storage in oxide matrices kept at 800-1000 deg. C will be discussed. In addition, results of He-implantation studies for metal matrices (W, Mo, Nb and V alloys) will be presented. Introduction of helium in metals at elevated temperatures leads to clustering of helium to bubbles. When operational temperatures are higher than 0.5 melting temperature, swelling and helium embrittlement might occur

  18. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  19. Helium accumulation and bubble formation in FeCoNiCr alloy under high fluence He+ implantation

    Science.gov (United States)

    Chen, Da; Tong, Y.; Li, H.; Wang, J.; Zhao, Y. L.; Hu, Alice; Kai, J. J.

    2018-04-01

    Face-centered cubic (FCC) high-entropy alloys (HEA), as emerging alloys with equal-molar or near equal-molar constituents, show a promising radiation damage resistance under heavy ion bombardment, making them potential for structural material application in next-generation nuclear reactors, but the accumulation of light helium ions, a product of nuclear fission reaction, has not been studied. The present work experimentally studied the helium accumulation and bubble formation at implantation temperatures of 523 K, 573 K and 673 K in a homogenized FCC FeCoNiCr HEA, a HEA showing excellent radiation damage resistance under heavy ion irradiation. The size and population density of helium bubbles in FeCoNiCr samples were quantitatively analyzed through transmission electron microscopy (TEM), and the helium content existing in bubbles were estimated from a high-pressure Equation of State (EOS). We found that the helium diffusion in such condition was dominated by the self-interstitial/He replacement mechanism, and the corresponding activation energy in FeCoNiCr is comparable with the vacancy migration energy in Ni and austenitic stainless steel but only 14.3%, 31.4% and 51.4% of the accumulated helium precipitated into helium bubbles at 523 K, 573 K and 673 K, respectively, smaller than the pure Ni case. Importantly, the small bubble size suggested that FeCoNiCr HEA has a high resistance of helium bubble formation compared with Ni and steels.

  20. Report of the reactor Operators Service - Annex F

    International Nuclear Information System (INIS)

    Zivotic, Z.

    1992-01-01

    RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. For continuous training of the existing operator staff the Service has prepared and published eleven booklets: Nuclear reactor; RA reactor primary coolant loop; System for purification of heavy water; reactor helium system; system for technical water; electric power system; control and operation; ventilation system in the reactor building; special sewage system; construction properties of the reactor core; reactor building and installations. During the reporting period there have been no accidents nor incidents that could affect the reactor personnel [sr

  1. Cryopumping of deuterium hydrogen and helium mixtures on smooth 4.2 K surfaces

    International Nuclear Information System (INIS)

    Chou, T.S.; Halama, H.J.

    1977-01-01

    The large quantities of deuterium and hydrogen to be pumped in a fusion reactor and its subsystems favor cryopumping over other pumping methods. Cryogen consumption and the operating pressure will not only depend on the gas to be pumped, but also on the amount of helium gas present in the system. In fact, residual helium pressure between pulses will determine the power dissipation of the pump, and hence, influence the choice of cryocondensation or cryosorption. In this paper we will present the results of our studies on (1) cryotrapping of helium in thick D 2 and H 2 films and hydrogen in D 2 films at 4.2 K; (2) diffusion of He and H 2 from D 2 films; (3) steady-state liquid helium consumption; (4) liquid helium consumption as a function of D 2 and H 2 fluxes being pumped; (5) liquid helium consumption as a function of He partial pressure in the system. Finally, these measurements will suggest maximum permissible He to D 2 and He to H 2 ratios in a cryocondensation pump

  2. Properties of vanadium-base alloys irradiated in the Dynamic Helium Charging Experiment*1

    Science.gov (United States)

    Chung, H. M.; Loomis, B. A.; Smith, D. L.

    1996-10-01

    One property of vanadium-base alloys that is not well understood in terms of their potential use a fusion reactor structural materials, is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of ≈ 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18-31 dpa at 425-600°C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V5Ti, V3Ti1Si, V8Cr6Ti, and V4Cr4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420°C. However, postirradiation ductilities at irradiation. Ductile—brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > - 150°C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at - 196°C in some specimens that were irradiated to 31 dpa at 425°C during the DHCE. For the helium generation rates in this experiment (≈ 0.4-4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

  3. Russian-American venture designs new reactor

    International Nuclear Information System (INIS)

    Newman, P.

    1994-01-01

    Russian and American nuclear energy experts have completed a joint design study of a small, low-cost and demonstrably accident-proof reactor that they say could revolutionize the way conventional reactors are designed, marketed and operated. The joint design is helium-cooled and graphite-moderated and has a power density of 3 MWt/cubic meter, which is significantly less than the standard American reactor. A prototype of this design should be operating in Chelyabinsk by June 1996

  4. Helium diffusion in nickel at high temperatures

    International Nuclear Information System (INIS)

    Philipps, V.

    1980-09-01

    Helium has been implanted at certain temperatures between 800 and 1250 0 C into single and polycrystalline Ni-samples with implantation depths between 15 and 90 μm. Simultaneously the helium reemission from the sample is measured by a mass-spectrometer. It has been shown that the time dependence of the observed reemission rate is governed by volume diffusion of the helium. Measuring this time dependence as a function of temperature the helium diffusion constant has been determined. The He-diffusion is interpreted as a interstitial diffusion hindered by thermal vacancies. Depending on the implantation depth more or less of the implanted helium remains in the sample and forms large helium bubbles. (orig./GSCH)

  5. Test of a cryogenic helium pump

    International Nuclear Information System (INIS)

    Lue, J.W.; Miller, J.R.; Walstrom, P.L.; Herz, W.

    1981-01-01

    The design of a cryogenic helium pump for circulating liquid helium in a magnet and the design of a test loop for measuring the pump performance in terms of mass flow vs pump head at various pump speeds are described. A commercial cryogenic helium pump was tested successfully. Despite flaws in the demountable connections, the piston pump itself has performed satisfactorily. A helium pump of this type is suitable for the use of flowing supercritical helium through Internally Cooled Superconductor (ICS) magnets. It has pumped supercritical helium up to 7.5 atm with a pump head up to 2.8 atm. The maximum mass flow rate obtained was about 16 g/s. Performance of the pump was degraded at lower pumping speeds

  6. Prospects of power conversion technology of direct-cycle helium gas turbine for MHTGR

    International Nuclear Information System (INIS)

    Li Yong; Zhang Zuoyi

    1999-01-01

    The modular high temperature gas cooled reactor (MHTGR) is a modern passively safe reactor. The reactor and helium gas turbine may be combined for high efficiency's power conversion, because MHTGR has high outlet temperature up to 950 degree C. Two different schemes are planed separately by USA and South Africa. the helium gas turbine methodologies adopted by them are mainly based on the developed heavy duty industrial and aviation gas turbine technology. The author introduces the differences of two technologies and some design issues in the design and manufacture. Moreover, the author conclude that directly coupling a closed Brayton cycle gas turbine concept to the passively safe MHTGR is the developing direction of MHTGR due to its efficiency which is much higher than that of using steam turbine

  7. Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

    Energy Technology Data Exchange (ETDEWEB)

    Crone, Wendy; Cao, Guoping; Sridhara, Kumar

    2013-05-31

    The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H{sub 2}, H{sub 2}O, CH{sub 4}, CO, CO{sub 2}, and N{sub 2} from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures. The project team will: • Evaluate stress-assisted grain boundary oxidation and creep crack initiation and crack growth in the temperature range of 500-850°C in a prototypical helium environment. • Evaluate the effects of oxygen partial pressure on stress-assisted grain boundary oxidation and creep crack growth in impure helium at 500°C, 700°C, and 850°C respectively. • Characterize the microstructure of candidate alloys after long-term exposure to an impure helium environment in order to understand the correlation between stress-assisted grain boundary oxidation, creep crack growth, material composition, and impurities in the helium coolant. • Evaluate grain boundary engineering as a method to mitigate stress-assisted grain boundary oxidation and creep crack growth of candidate alloys in impure helium. The maximum primary helium coolant temperature in the high-temperature reactor is expected to be 850-1,000°C.Corrosion may involve oxidation, carburization, or decarburization mechanisms depending on the temperature, oxygen partial pressure, carbon activity, and alloy composition. These corrosion reactions can substantially affect long-term mechanical properties such as crack- growth rate and fracture

  8. Numerical Study on the Helium Flow Characteristics for Steam Generator Subsystem of HTR

    International Nuclear Information System (INIS)

    Ha, Jung Hoon; Ham, Jin Ki; Ki, Min-Hwan; Lee, Won Jae

    2014-01-01

    The High Temperature Reactor (HTR), one of the 4th generation reactors, utilizes helium as the primary coolant. A Steam Generator Subsystem (SGS) is installed to transfer heat from the primary coolant to feed water and subsequently produce steam so that it supplies electricity as well as process heat over a wide range. The SGS is composed of a helical heat exchanger, shrouds directing the flow of the shell side helium and support systems, which are located within the steam generator vessel. In this study, helium flow characteristics in the SGS were investigated at various operating conditions using Computational Fluid Dynamics (CFD). A full-scale 3-D model of the SGS was developed and the reynolds stress model with standard wall treatment was used as a turbulence model. The CFD result was compared to that of the concept design of the steam cycle modular helium reactor for the design verification of the SGS. From the CFD analysis, it was found that the primary coolant flow had non-uniform distribution while it passed the inlet in the helical heat exchanger. In order to make the uniform primary coolant flow uniform, a special type of screen was suggested in front of the helical heat exchanger. As a result, the overall design adequacy of the SGS has been evaluated. (author)

  9. Backscattered Helium Spectroscopy in the Helium Ion Microscope: Principles, Resolution and Applications

    NARCIS (Netherlands)

    van Gastel, Raoul; Hlawacek, G.; Dutta, S.; Poelsema, Bene

    2015-01-01

    We demonstrate the possibilities and limitations for microstructure characterization using backscattered particles from a sharply focused helium ion beam. The interaction of helium ions with matter enables the imaging, spectroscopic characterization, as well as the nanometer scale modification of

  10. Desorption of tritium and helium from high dose neutron irradiated beryllium

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

    2007-08-01

    The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

  11. Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

    International Nuclear Information System (INIS)

    He, Qingyun; Feng, Jingchao; Chen, Hongli

    2016-01-01

    Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

  12. Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

    Energy Technology Data Exchange (ETDEWEB)

    He, Qingyun; Feng, Jingchao; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

    2016-02-15

    Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

  13. Helium leak testing methods in nuclear applications

    International Nuclear Information System (INIS)

    Ahmad, Anis

    2004-01-01

    Helium mass-spectrometer leak test is the most sensitive leak test method. It gives very reliable and sensitive test results. In last few years application of helium leak testing has gained more importance due to increased public awareness of safety and environment pollution caused by number of growing chemical and other such industries. Helium leak testing is carried out and specified in most of the critical area applications like nuclear, space, chemical and petrochemical industries

  14. Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

    Energy Technology Data Exchange (ETDEWEB)

    Chu, X. X.; Zhang, D. X.; Qian, Y.; Liu, W. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai, 201800 (China); Zhang, M. M.; Xu, D. [Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing, 100190 (China)

    2014-01-29

    In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in helium recycle gas are less than 1 ppb.

  15. Behaviour of helium after implantation in molybdenum

    International Nuclear Information System (INIS)

    Viaud, C.; Maillard, S.; Carlot, G.; Valot, C.; Gilabert, E.; Sauvage, T.; Peaucelle, C.; Moncoffre, N.

    2009-01-01

    This study deals with the behaviour of helium in a molybdenum liner dedicated to the retention of fission products. More precisely this work contributes to evaluate the release of implanted helium when the gas has precipitated into nanometric bubbles close to the free surface. A simple model dedicated to calculate the helium release in such a condition is presented. The specificity of this model lays on the assumption that the gas is in equilibrium with a simple distribution of growing bubbles. This effort is encouraging since the calculated helium release fits an experimental dataset with a set of parameters in good agreement with the literature

  16. Helium turbo-expander with an alternator

    International Nuclear Information System (INIS)

    Akiyama, Yoshitane

    1980-01-01

    Study was made on a helium turbo-expander, the heart of helium refrigerator systems, in order to develop a system which satisfies the required conditions. A helium turbo-expander with externally pressurized helium gas bearings at the temperature of liquid nitrogen and an alternator as a brake have been employed. The essential difference between a helium turbo-expander and a nitrogen turbo-expander was clarified. The gas bearing lubricated with nitrogen at room temperature and the gas bearing lubricated with helium at low temperature were tested. The flow rate of helium in a helium refrigerator for a large superconducting magnet is comparatively small, therefore a helium turbine must be small, but the standard for large turbine design can be applied to such small turbine. Using the alternator as a brake, the turbo-expander was easily controllable electrically. The prototype turbo-expander was made, and the liquefaction test with it and MHD power generation test were carried out. (Kako, I.)

  17. Helium Extraction from LNG End Flash

    OpenAIRE

    Kim, Donghoi

    2014-01-01

    Helium is an invaluable element as it is widely used in industry such as cryo-genics and welding due to its unique properties. However, helium shortage is expected in near future because of increasing demand and the anxiety of sup-ply. Consequently, helium production has attracted the attention of industry. The main source of He is natural gas and extracting it from LNG end-flash is considered as the most promising way of producing crude helium. Thus, many process suppliers have proposed proc...

  18. Helium supply demand in future years

    International Nuclear Information System (INIS)

    Laverick, C.

    1975-01-01

    Adequate helium will be available to the year 2000 AD to meet anticipated helium demands for present day applications and the development of new superconducting technologies of potential importance to the nation. It is almost certain that there will not be enough helium at acceptable financial and energy cost after the turn of the century to meet the needs of the many promising helium based technologies now under development. Serious consideration should be given to establishing priorities in development and application based upon their relative value to the country. In the first half of the next century, three ways of estimating helium demand lead to cumulative ranges of from 75 to 125 Gcf (economic study), 89 to 470 Gcf (projected national energy growth rates) and 154 to 328 Gcf (needs for new technologies). These needs contrast with estimated helium resources in natural gas after 2000 AD which may be as low as 10 or 126 Gcf depending upon how the federal helium program is managed and the nation's natural gas resources are utilized. The technological and financial return on a modest national investment in further helium storage and a rational long term helium program promises to be considerable

  19. Fast leak of a channel filled with helium at a pressure of 2 bars (channel H5)

    International Nuclear Information System (INIS)

    Bauer, E.; Tribolet, J.

    1987-01-01

    The loss of seal of a helium-filled channel opening the entire cross section of the front part leads to a fast leak. The channel fills to the upper generatrix of the leak orifice and part of the helium contained in the channel escapes into the circuit. The pressure drop in the reflector can lead to reactor and main pump shutdown. On the other hand, the Cooling Circuit Shutdown Bar circuit pumps remain in operation. This paper evaluates the consequences of an incident of this nature for the reactor and the surrounding experimental zones

  20. The assessment of helium purification system of small power HTGR

    International Nuclear Information System (INIS)

    Siti Alimah; Sriyono

    2016-01-01

    The helium purification system (HPS) is one of safety system of High Temperature Gas-cooled Reactor. HPS removes impurities in the primary coolant, so that the impact on structure, system and component (SSC) is minimized. The two impurity types are particulates (carbon dust, fission products (Kr, Xe, Cs etc.) and the gases (O_2, N_2, H_2O, CH_4, CO, CO_2 and H_2). Every reactor has a different impurity limit during normal operation, depends on the reactor power, energy conversion system and fuel type. This paper discusses the HPS on HTR-10, HTTR and Indonesian RDE conceptual design. The purpose of this assessment is to determine the optimum HPS design as a role model for Indonesian RDE. The utilized methodology is a literature study based on the operating experiences of both HTR-10 and HTTR as well as the evaluation of RDE conceptual design. This study focuses on the impurities limit during normal operation, the main components of HPS, mass flow-rate and regeneration process. The main component that used in HPS for HTR-10, HTTR and RDE are similar i.e. filter, CuO column, water cooler, molecular sieve bed and cryogenic activated carbon bed. Refer to the HTR-10 and HTTR operational experiences, both of those reactors have a purification systems that capable to maintain the helium purity, even though the impurities limit are different. The HPS of HTTR Japan has a stricter impurities limit that N_2, CH_4, and O_2 should not be contained at all during normal operation and the pre-charcoal trap is used to adsorb the fine dust below 0.1 micron. Both of these parameters can be adopted to the RDE's HPS design to minimize the effect of impurities to SSC. (author)

  1. New helium spectrum variable and a new helium-rich star

    International Nuclear Information System (INIS)

    Walborn, N.R.

    1974-01-01

    HD 184927, known previously as a helium-rich star, has been found to have a variable helium spectrum; the equivalent widths of five He I lines are larger by an average of 46 percent on a 1974 spectrogram than on one obtained with the same equipment in 1970. HD 186205 has been found to be a new, pronounced helium-rich star. (auth)

  2. Helium transfer line installation details.

    CERN Multimedia

    G. Perinic

    2007-01-01

    A particularity of the 32 m long four in one helium transfer line in between the cold box in USC55 and the cavern UX5 is the fact that the transfer line passes through a hole in the crane rail support beam. In order to ensure the alignment of the suspension rail in the interconnecting tunnel with the hole in the rail support as well as the connection points at both ends required precise measurements of the given geometries as well as the installation of a temporary target for the verification of the theoretical predictions.

  3. Helium bubble bursting in tungsten

    International Nuclear Information System (INIS)

    Sefta, Faiza; Juslin, Niklas; Wirth, Brian D.

    2013-01-01

    Molecular dynamics simulations have been used to systematically study the pressure evolution and bursting behavior of sub-surface helium bubbles and the resulting tungsten surface morphology. This study specifically investigates how bubble shape and size, temperature, tungsten surface orientation, and ligament thickness above the bubble influence bubble stability and surface evolution. The tungsten surface is roughened by a combination of adatom “islands,” craters, and pinholes. The present study provides insight into the mechanisms and conditions leading to various tungsten topology changes, which we believe are the initial stages of surface evolution leading to the formation of nanoscale fuzz

  4. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

    International Nuclear Information System (INIS)

    Beech, D.J.; May, R.

    2000-01-01

    In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

  5. Spectroscopy of antiproton helium atoms

    International Nuclear Information System (INIS)

    Hayano, Ryugo

    2005-01-01

    Antiproton helium atom is three-body system consisting of an antiproton, electrons and a helium nucleus (denoted by the chemical symbol, p-bar H + ). The authors produced abundant atoms of p-bar 4 He + , and p-bar 3 He + in a cooled He gas target chamber stopping the p-bar beam decelerated to approximately 100 keV in the Antiproton Decelerator at CERN. A precision laser spectroscopy on the atomic transitions in the p-bar 4 He + , and in p-bar 3 He + was performed. Principle of laser spectroscopy and various modifications of the system to eliminate factors affecting the accuracy of the experiment were described. Deduced mass ratio of antiproton and proton, (|m p -bar - m p |)/m p reached to the accuracy of 10 ppb (10 -8 ) as of 2002, as adopted in the recent article of the Particle Data Group by P.J. Mohr and B.N. Taylor. This value is the highest precise data for the CPT invariance in baryon. In future, antihydrogen atoms will be produced in the same facility, and will provide far accurate value of antiproton mass thus enabling a better confirmation of CPT theorem in baryon. (T. Tamura)

  6. Experimental method to determine the role of helium in neutron-induced microstructural evolution

    International Nuclear Information System (INIS)

    Gelles, D.S.; Garner, F.A.

    1978-12-01

    A method is presented which allows the determination of the role of helium on microstructural evolution in complex alloys and which avoids many of the problems associated with other simulation experiments. It involves a direct comparison of the materials' response to a primary difference in fission and fusion environments, namely the rate of helium generation. This is accomplished by irradiating specimens in a fission reactor and conducting microstructural analyses which concentrate on alloy matrix regions adjacent to precipitates rich in boron or nitrogen. Procedures are outlined for calculation of background and injected helium levels as well as displacement doses generated by neutrons and alpha particles. An example of the analysis method is shown for an experimental austenitic stainless steel containing boride particles and irradiated to 3 and 7 x 10 22 n/cm 2

  7. Strange matter and Big Bang helium synthesis

    International Nuclear Information System (INIS)

    Madsen, J.; Riisager, K.

    1985-01-01

    Stable strange quark matter produced in the QCD phase transition in the early universe will trap neutrons and repel protons, thus reducing primordial helium production, Ysub(p). For reasonable values of Ysub(p), the radius of strange droplets must exceed 10 -6 cm if strange matter shall solve the dark-matter problem without spoiling Big Bang helium synthesis. (orig.)

  8. Post-giant evolution of helium stars

    International Nuclear Information System (INIS)

    Schoenberner, D.

    1977-01-01

    Extremely hydrogen deficient stars (helium stars and R Coronae Borealis variables) are considered to be remnants of double shell source stars (of the asymptotic giant branch). The evolution of stars with a condensed C/O-core and a helium envelope is followed numerically from the red giant stage to the white dwarf domain, crossing the regions of R CrB- and helium stars (so far analyzed). They have typically masses M/M(sun) = 0.7 and luminosities log L/L(sun) = 4.1. The time for crossing the helium star domain is some 10 3 years. The corresponding times in the R CrB-region amounts up to several 10 4 years. The lower limit of the death rate of helium stars is estimated to be 4 x 10 -14 pc -3 yr -1 . This value is only a factor of ten lower than the birth rate of all non-DA white dwarfs. It is therefore possible that the helium stars are the precursors of helium rich white dwarfs. As a consequence, a significant fraction of all stars which end their lives as white dwarfs should pass through the helium star phase. (orig.) [de

  9. Organ protection by the noble gas helium

    NARCIS (Netherlands)

    Smit, K.F.

    2017-01-01

    The aims of this thesis were to investigate whether helium induces preconditioning in humans, and to elucidate the mechanisms behind this possible protection. First, we collected data regarding organ protective effects of noble gases in general, and of helium in particular (chapters 1-3). In chapter

  10. Radioactive ions and atoms in superfluid helium

    NARCIS (Netherlands)

    Dendooven, P.G.; Purushothaman, S.; Gloos, K.; Aysto, J.; Takahashi, N.; Huang, W.; Harissopulos, S; Demetriou, P; Julin, R

    2006-01-01

    We are investigating the use of superfluid helium as a medium to handle and manipulate radioactive ions and atoms. Preliminary results on the extraction of positive ions from superfluid helium at temperatures close to 1 K are described. Increasing the electric field up to 1.2 kV/cm did not improve

  11. Modeling Secondary Neutral Helium in the Heliosphere

    International Nuclear Information System (INIS)

    Müller, Hans-Reinhard; Möbius, Eberhard; Wood, Brian E.

    2016-01-01

    An accurate, analytic heliospheric neutral test-particle code for helium atoms from the interstellar medium (ISM) is coupled to global heliospheric models dominated by hydrogen and protons from the solar wind and the ISM. This coupling enables the forward-calculation of secondary helium neutrals from first principles. Secondaries are produced predominantly in the outer heliosheath, upwind of the heliopause, by charge exchange of helium ions with neutral atoms. The forward model integrates the secondary production terms along neutral trajectories and calculates the combined neutral helium phase space density in the innermost heliosphere where it can be related to in-situ observations. The phase space density of the secondary component is lower than that of primary neutral helium, but its presence can change the analysis of primaries and the ISM, and can yield valuable insight into the characteristics of the plasma in the outer heliosheath. (paper)

  12. Exotic helium molecules; Molecules exotiques d'helium

    Energy Technology Data Exchange (ETDEWEB)

    Portier, M

    2007-12-15

    We study the photo-association of an ultracold cloud of magnetically trapped helium atoms: pairs of colliding atoms interact with one or two laser fields to produce a purely long range {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}P{sub 0}) molecule, or a {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}S{sub 1}) long range molecule. Light shifts in one photon photo-association spectra are measured and studied as a function of the laser polarization and intensity, and the vibrational state of the excited molecule. They result from the light-induced coupling between the excited molecule, and bound and scattering states of the interaction between two metastable atoms. Their analysis leads to the determination of the scattering length a = (7.2 {+-} 0.6) ruling collisions between spin polarized atoms. The two photon photo-association spectra show evidence of the production of polarized, long-range {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}S{sub 1}) molecules. They are said to be exotic as they are made of two metastable atoms, each one carrying a enough energy to ionize the other. The corresponding lineshapes are calculated and decomposed in sums and products of Breit-Wigner and Fano profiles associated to one and two photon processes. The experimental spectra are fit, and an intrinsic lifetime {tau} = (1.4 {+-} 0.3) {mu}s is deduced. It is checked whether this lifetime could be limited by spin-dipole induced Penning autoionization. This interpretation requires that there is a quasi-bound state close to the dissociation threshold in the singlet interaction potential between metastable helium atoms for the theory to match the experiment. (author)

  13. Exotic helium molecules; Molecules exotiques d'helium

    Energy Technology Data Exchange (ETDEWEB)

    Portier, M

    2007-12-15

    We study the photo-association of an ultracold cloud of magnetically trapped helium atoms: pairs of colliding atoms interact with one or two laser fields to produce a purely long range {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}P{sub 0}) molecule, or a {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}S{sub 1}) long range molecule. Light shifts in one photon photo-association spectra are measured and studied as a function of the laser polarization and intensity, and the vibrational state of the excited molecule. They result from the light-induced coupling between the excited molecule, and bound and scattering states of the interaction between two metastable atoms. Their analysis leads to the determination of the scattering length a = (7.2 {+-} 0.6) ruling collisions between spin polarized atoms. The two photon photo-association spectra show evidence of the production of polarized, long-range {sup 4}He{sub 2}(2{sup 3}S{sub 1}-2{sup 3}S{sub 1}) molecules. They are said to be exotic as they are made of two metastable atoms, each one carrying a enough energy to ionize the other. The corresponding lineshapes are calculated and decomposed in sums and products of Breit-Wigner and Fano profiles associated to one and two photon processes. The experimental spectra are fit, and an intrinsic lifetime {tau} = (1.4 {+-} 0.3) {mu}s is deduced. It is checked whether this lifetime could be limited by spin-dipole induced Penning autoionization. This interpretation requires that there is a quasi-bound state close to the dissociation threshold in the singlet interaction potential between metastable helium atoms for the theory to match the experiment. (author)

  14. The 1/3-scale aerodynamics performance test of helium compressor for GTHTR300 turbo machine of JAERI (step 1)

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Xing, Yan

    2003-01-01

    A program for research and development on aerodynamics in a helium gas compressor was planned for the power conversion system of the Gas Turbine High Temperature Reactor (GTHTR300). The three-dimensional aerodynamic design of the compressor achieved a high polytropic efficiency of 90%, keeping a sufficient surge margin over 30%. To validate the design of the helium gas compressor of GTHTR300, aerodynamic performance tests were planned, and a 1/3-scale, 4-stage compressor model was designed. In the tests, the performance data of the helium gas compressor model will be acquired by using helium gas as a working fluid. The maximum design pressure at the model inlet is 0.88 MPa, which allows the Reynolds number to be sufficiently high. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  15. Studies on MHD pressure drop and heat transfer of helium-lithium annular-mist flow in a transverse magnetic field

    International Nuclear Information System (INIS)

    Inoue, Akira; Aritomi, Masanori; Takahashi, Minoru; Matsuzaki, Mitsuo; Narita, Yoshihito; Yano, Toshikazu.

    1987-01-01

    Pressure drop and heat transfer coefficient of helium-lithium annular-mist flow in a rectangular duct were investigated experimentally under a transverse magnetic field at system pressure of 0.2 MPa. A ratio of MHD pressure drop to that of non-magnetic field increases with magnetic flux density and a mass flow rate ratio of lithium to helium in low helium velocity region. However, as increasing the helium velocity, the increment of MHD pressure drop with the magnetic flux density is much reduced and then becomes almost zero. At this condition, the MHD pressure drop of the annular-mist flow becomes much smaller than that of lithium single phase flow with the same lithium mass flow at the high magnetic flux density. Heat transfer coefficient ratio of the helium-lithium annular-mist flow to helium single phase in the non-magnetic field is well correlated by a ratio of the mass flow rate of lithium to helium. The heat transfer coefficient in the magnetic field increases with the magnetic flux density and then terminates at a certain value depending on the mass flow rate ratio and the helium velocity. These characteristics of the MHD pressure drop and the heat transfer in the magnetic field suggest that the helium-lithium annular-mist flow is effectively applicable to cooling of the high heat flux wall in a strong magnetic field like a first wall of a magnetic confinement fusion reactors. (author)

  16. Determination of helium in beryl minerals

    International Nuclear Information System (INIS)

    Souza Barcellos, E. de.

    1985-08-01

    In order to obtain the diffusion coefficients of helium in beryl and phenacite samples at various temperatures, helium leak rates were measured in these minerals at these temperatures. Mass spectrometry (MS) was used to obtain helium leak rates and the gas flow was plotted against time. The gas quantity determined by MS was first obtained at various temperatures until no helium leak rate was detected. After that, these samples were irradiated with fast neutrons to produce helium which was measured again. This procedure was used to estimate the experimental error. The quantity of helium produced by interaction of gamma radiation with beryl minerals was theoretically calculated from the amount of thorium-232 at the neighbourhood of the samples. The quantity of helium produced in the minerals due to uranium and thorium decay was calculated using the amount of these heavy elements, and the results were compared with the amounts determined by MS. The amount of potassium-40 was determined in order to derive the quantity of argonium-40, since some workers found argonium in excess in these minerals. The quantity of helium in the beryl samples (s) was determined in the center and in the surface of the samples in order to obtain informations about the effectiveness of the Be(α, η) He reaction. Beryl and phenacite minerals were choosen in this research since they are opposite each other with respect to the helium contents. Both have beryllium in their compositon but beryl hold a large amount of helium while phenacite, in spite of having about three times more beryllium than beryl, do not hold the gas. (author) [pt

  17. Commercial helium reserves, continental rifting and volcanism

    Science.gov (United States)

    Ballentine, C. J.; Barry, P. H.; Hillegonds, D.; Fontijn, K.; Bluett, J.; Abraham-James, T.; Danabalan, D.; Gluyas, J.; Brennwald, M. S.; Pluess, B.; Seneshens, D.; Sherwood Lollar, B.

    2017-12-01

    Helium has many industrial applications, but notably provides the unique cooling medium for superconducting magnets in medical MRI scanners and high energy beam lines. In 2013 the global supply chainfailed to meet demand causing significant concern - the `Liquid Helium Crisis' [1]. The 2017 closure of Quatar borders, a major helium supplier, is likely to further disrupt helium supply, and accentuates the urgent need to diversify supply. Helium is found in very few natural gas reservoirs that have focused 4He produced by the dispersed decay (a-particle) of U and Th in the crust. We show here, using the example of the Rukwa section of the Tanzanian East African Rift, how continental rifting and local volcanism provides the combination of processes required to generate helium reserves. The ancient continental crust provides the source of 4He. Rifting and associated magmatism provides the tectonic and thermal mechanism to mobilise deep fluid circulation, focusing flow to the near surface along major basement faults. Helium-rich springs in the Tanzanian Great Rift Valley were first identified in the 1950's[2]. The isotopic compositions and major element chemistry of the gases from springs and seeps are consistent with their release from the crystalline basement during rifting [3]. Within the Rukwa Rift Valley, helium seeps occur in the vicinity of trapping structures that have the potential to store significant reserves of helium [3]. Soil gas surveys over 6 prospective trapping structures (1m depth, n=1486) show helium anomalies in 5 out of the 6 at levels similar to those observed over a known helium-rich gas reservoir at 1200m depth (7% He - Harley Dome, Utah). Detailed macroseep gas compositions collected over two days (n=17) at one site allows us to distinguish shallow gas contributions and shows the deep gas to contain between 8-10% helium, significantly increasing resource estimates based on uncorrected values (1.8-4.2%)[2,3]. The remainder of the deep gas is

  18. Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

    International Nuclear Information System (INIS)

    Rabaglino, E.; Ronchi, C.; Cardella, A.

    2003-01-01

    One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release

  19. Mobility of hydrogen-helium clusters in tungsten studied by molecular dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Grigorev, Petr, E-mail: grigorievpit@gmail.com [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St.Petersburg Polytechnic University, St. Petersburg (Russian Federation); Terentyev, Dmitry; Bonny, Giovanni [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Zhurkin, Evgeny E. [Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St.Petersburg Polytechnic University, St. Petersburg (Russian Federation); Oost, Guido van [Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Noterdaeme, Jean-Marie [Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany)

    2016-06-15

    Tungsten is a primary candidate material for plasma facing components in fusion reactors. Interaction of plasma components with the material is unavoidable and will lead to degradation of the performance and the lifetime of the in-vessel components. In order to gain better understanding the mechanisms driving the material degradation at atomic level, atomistic simulations are employed. In this work we study migration, stability and self-trapping properties of pure helium and mixed helium-hydrogen clusters in tungsten by means of molecular dynamics simulations. We test two versions of an embedded atom model interatomic potential by comparing it with ab initio data regarding the binding properties of He clusters. By analysing the trajectories of the clusters during molecular dynamics simulations at finite temperatures we obtain the diffusion parameters. The results show that the diffusivity of mixed clusters is significantly lower, than that of pure helium clusters. The latter suggest that the formation of mixed clusters during mixed hydrogen helium plasma exposure will affect the helium diffusivity in the material.

  20. Helium hammer in superfluid transfer

    Science.gov (United States)

    Tward, E.; Mason, P. V.

    1984-01-01

    Large transient pressure pulses, referred to as a helium hammer, which occurred in the transfer line of the main cryogenic tank during the development tests of the Infrared Astronomical Satellite, launched on January 25, 1983, are analyzed, and the measures taken to prevent a failure described. The modifications include an installation of a 2.3-liter surge tank upstream, and a back-up relief valve downstream, of a burst disk. The surge tank is designed to attenuate a 0.33-MPa pressure pulse at the inlet down to 0.092 MPa at the outlet. A mechanism of the pulse generation is suggested, which involves flashing and rapid recondensation of the small amount of liquid entering the warm section of a transition to room temperature.

  1. Friction and wear studies of graphite and a carbon-carbon composite in air and in helium

    International Nuclear Information System (INIS)

    Li, C.C.; Sheehan, J.E.

    1980-10-01

    Sliding friction and wear tests were conducted on a commercial isotropic graphite and a carbon-carbon composite in air, purified helium, and a helium environment containing controlled amounts of impurities simulating the primary coolant chemistry of a high-temperature gas-cooled reactor (HTGR). The friction and wear characteristics of the materials investigated were stable and were found to be very sensitive to the testing temperature. In general, friction and wear decreased with increasing temperature in the range from ambient to 950 0 C. This temperature dependence is concluded to be due to chemisorption of impurities to form lubricating films and oxidation at higher temperatures, which reduce friction and wear. Graphite and carbon-carbon composites are concluded to be favorable candidate materials for high-temperature sliding service in helium-cooled reactors

  2. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Golay, M.W.

    1990-01-01

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  3. Friendly fermions of helium-three

    International Nuclear Information System (INIS)

    Leggatt, T.

    1976-01-01

    The importance of helium in showing up the effects of atomic indistinguishability and as a material by which to test some of the most fundamental principles of quantum mechanics is discussed. Helium not only remains liquid down to zero temperature but of the two isotopes helium-three has intrinsic spin 1/2 and should therefore obey the Pauli principle, while helium-four has spin zero and is expected to undergo Bose condensation. Helium-three becomes superfluid at temperatures of a few thousandths of a degree above absolute zero by the bulk liquid collecting its atoms into spinning pairs. There are three different superfluid phases, now conveniently called A, B and A 1 and each is characterised by a different behaviour of the spin and/or relative angular motion of the atoms composing the Cooper pairs. Problems surrounding the complicated physical system of helium-three are discussed. It is suggested that the combined coherence and directionality of superfluid helium-three should create some fascinating physics. (U.K.)

  4. Seismological measurement of solar helium abundance

    International Nuclear Information System (INIS)

    Vorontsov, S.V.; Pamyatnykh, A.A.

    1991-01-01

    The internal structure and evolution of the Sun depends on its chemical composition, particularly the helium abundance. In addition, the helium abundance in the solar envelope is thought to represent the protosolar value, making it a datum of cosmological significance. Spectroscopic measurements of the helium abundance are uncertain, and the most reliable estimates until now have come from the calibration of solar evolutionary models. The frequencies of solar acoustic oscillations are sensitive, however, to the behaviour of the speed of sound in the Sun's helium ionization zone, which allows a helioseismological determination of the helium abundance. Sound-speed inversion of helioseismological data can be used for this purpose, but precise frequency measurements of high-degree oscillation modes are needed. Here we describe a new approach based on an analysis of the phase shift of acoustic waves of intermediate-degree modes. From the accurate intermediate-mode data now available, we obtain a helium mass fraction Y=0.25±0.01 in the solar convection zone, significantly smaller than the value Y=0.27-0.29 predicted by recent solar evolutionary models. The discrepancy indicates either that initial helium abundance was reduced in the envelope by downward diffusion or that the protosolar value was lower than currently accepted. (author)

  5. Report on the control of the safety and security of nuclear facilities. Part 2: the reconversion of military plutonium stocks. The use of the helps given to central and eastern Europe countries and to the new independent states; Rapport sur le controle de la surete et de la securite des installations nucleaires. Deuxieme partie: la reconversion des stocks de plutonium militaire. L'utilisation des aides accordees aux pays d'Europe centrale et orientale et aux nouveaux etats independants

    Energy Technology Data Exchange (ETDEWEB)

    Birraux, C

    2002-07-01

    This report deals with two different aspects of the safety and security of nuclear facilities. The first aspect concerns the reconversion of weapon grade plutonium stocks: the plutonium in excess, plutonium hazards and nuclear fuel potentialities, the US program, the Russian program, the actions of European countries (France, Germany), the intervention of other countries, the unanswered questions (political aspects, uncertainties), the solutions of the future (improvement of reactors, the helium-cooled high temperature reactor technology (gas-turbine modular helium reactor: GT-MHR), the Carlo Rubbia's project). The second aspect concerns the actions carried out by the European Union in favor of the civil nuclear facilities of central and eastern Europe: the European Union competencies through the Euratom treaty, the conclusions of the European audit office about the PHARE and TACIS nuclear programs, the status of committed actions, the coming planned actions, and the critical analysis of the policy adopted so far. (J.S.)

  6. Report on the control of the safety and security of nuclear facilities. Part 2: the reconversion of military plutonium stocks. The use of the helps given to central and eastern Europe countries and to the new independent states; Rapport sur le controle de la surete et de la securite des installations nucleaires. Deuxieme partie: la reconversion des stocks de plutonium militaire. L'utilisation des aides accordees aux pays d'Europe centrale et orientale et aux nouveaux etats independants

    Energy Technology Data Exchange (ETDEWEB)

    Birraux, C

    2002-07-01

    This report deals with two different aspects of the safety and security of nuclear facilities. The first aspect concerns the reconversion of weapon grade plutonium stocks: the plutonium in excess, plutonium hazards and nuclear fuel potentialities, the US program, the Russian program, the actions of European countries (France, Germany), the intervention of other countries, the unanswered questions (political aspects, uncertainties), the solutions of the future (improvement of reactors, the helium-cooled high temperature reactor technology (gas-turbine modular helium reactor: GT-MHR), the Carlo Rubbia's project). The second aspect concerns the actions carried out by the European Union in favor of the civil nuclear facilities of central and eastern Europe: the European Union competencies through the Euratom treaty, the conclusions of the European audit office about the PHARE and TACIS nuclear programs, the status of committed actions, the coming planned actions, and the critical analysis of the policy adopted so far. (J.S.)

  7. Report on the control of the safety and security of nuclear facilities. Part 2: the reconversion of military plutonium stocks. The use of the helps given to central and eastern Europe countries and to the new independent states

    International Nuclear Information System (INIS)

    Birraux, C.

    2002-01-01

    This report deals with two different aspects of the safety and security of nuclear facilities. The first aspect concerns the reconversion of weapon grade plutonium stocks: the plutonium in excess, plutonium hazards and nuclear fuel potentialities, the US program, the Russian program, the actions of European countries (France, Germany), the intervention of other countries, the unanswered questions (political aspects, uncertainties), the solutions of the future (improvement of reactors, the helium-cooled high temperature reactor technology (gas-turbine modular helium reactor: GT-MHR), the Carlo Rubbia's project). The second aspect concerns the actions carried out by the European Union in favor of the civil nuclear facilities of central and eastern Europe: the European Union competencies through the Euratom treaty, the conclusions of the European audit office about the PHARE and TACIS nuclear programs, the status of committed actions, the coming planned actions, and the critical analysis of the policy adopted so far. (J.S.)

  8. Superfluid helium-4: An introductory review

    International Nuclear Information System (INIS)

    Vinen, W.F.

    1983-01-01

    Helium was first liquefied by Kamerlingh Onnes in Leiden in July 1908, an achievement that followed much careful and painstaking work. On the same day Onnes reduced the temperature of his helium to a value approaching lK, and he must therefore have produced and observed the superfluid phase. These experimental discoveries led very quickly to a series of remarkable theoretical contributions that laid the foundations for all subsequent work. The period since the second world war has of course seen an enormous amount of work on superfluid helium-4. In reviewing it the author tries to see it in terms of two threads: one originating from Landau; the other from London

  9. Low-temperature centrifugal helium compressor

    International Nuclear Information System (INIS)

    Kawada, M.; Togo, S.; Akiyama, Y.; Wada, R.

    1974-01-01

    A centrifugal helium compressor with gas bearings, which can be operated at the temperature of liquid nitrogen, has been investigated. This compressor has the advantages that the compression ratio should be higher than the room temperature operation and that the contamination of helium could be eliminated. The outer diameter of the rotor is 112 mm. The experimental result for helium gas at low temperature shows a flow rate of 47 g/s and a compression ratio of 1.2 when the inlet pressure was 1 ata and the rotational speed 550 rev/s. The investigation is now focused on obtaining a compression ratio of 1.5. (author)

  10. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    International Nuclear Information System (INIS)

    Marvin, M.D.

    1978-01-01

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated

  11. 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

  12. Adsorption of krypton from helium by low temperature charcoal

    International Nuclear Information System (INIS)

    Cooper, M.H.; Simmons, C.R.; Taylor, G.R.

    1975-01-01

    Adsorption of krypton from helium by charcoal at temperatures from -100 0 C to -140 0 C was experimentally investigated to verify adsorption system design methods and to determine effects of regeneration for the Gas Purification System of the Liquid-Metal Fast Breeder Reactor. Helium with two krypton concentrations, traced by krypton-85 at 0.0044 μCi/cm 3 , was passed through a 1/2-inch diameter, three-inch long trap packed with coconut charcoal. Breakthrough curves were measured by continuously recording the activity of the effluent gas using a sampler with a krypton-85 detection limit of about 5 x 10 -7 μCi/cm 3 . Experimental breakthrough curves with continuous feed for both concentrations and for superficial gas velocities of 5 to 28 cm/sec were closely fitted when the pore diffusion term was omitted from the Anzelius linear equilibrium adsorption model indicating that the adsorption process for this system was controlled by gas phase mass transport kinetics. Adsorption capacities determined in these experiments at -140 0 C agreed closely with published data. A discontinuity, however, was observed in the krypton adsorption coefficient between -100 and -120 0 C. This discontinuity may be caused by capillary condensation of krypton in the charcoal pores. Breakthrough times for pulse experiments at 400 ppM (vol.) krypton concentration were several times greater than breakthrough for continuous feed experiments at equivalent conditions. The differences in breakthrough times indicate that the adsorption isotherms are non-linear in this concentration range. Regeneration experiments showed that purging with helium at room temperature for 16 hours was inadequate, since lower breakthrough times were obtained after this treatment. Regeneration under vacuum at 100 0 C or 200 0 C for 16 hours resulted in satisfactory regeneration (i.e., no reduction in breakthrough times occurred in subsequent runs). (U.S.)

  13. A study of formation of iron nanoparticles in aluminium matrix with helium pores

    Czech Academy of Sciences Publication Activity Database

    Kichanov, S.E.; Kozlenko, D. P.; Belushkin, A.V.; Reutov, V.F.; Samoilenko, S.O.; Jirák, Zdeněk; Savenko, B. N.; Bulavin, L. A.; Zubavichus, Y.V.

    2012-01-01

    Roč. 351, č. 1 (2012), "012013-1"-"012013-5" ISSN 1742-6588. [International Workshop on SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor - Devoted to the 75th anniversary of Yu M Ostanevich's Birth /2./. Dubna, 27.05.2011-30.05.2011] Institutional research plan: CEZ:AV0Z10100521 Keywords : iron nanoparticles * aluminium matrix * helium pores Subject RIV: BM - Solid Matter Physics ; Magnetism

  14. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  15. Fatigue behavior of Type 316 stainless steel following neutron irradiation inducing helium

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Liu, K.C.

    1980-01-01

    Since a tokamak reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied

  16. The primary circuit of the dragon high temperature reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.

    2005-01-01

    The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

  17. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  18. Use of separating nozzles or ultra-centrifuges for obtaining helium from gas mixtures containing helium

    International Nuclear Information System (INIS)

    Reimann, T.

    1987-01-01

    To obtain helium from gas mixtures containing helium, particularly from natural gas, it is proposed to match the dimensions of the separation devices for a ratio of the molecular weights to be separated of 4:1 of more, which ensures a higher separation factor and therefore a smaller number of separation stages to be connected in series. The process should make reasonably priced separation of helium possible. (orig./HP) [de

  19. Perspectives on Lunar Helium-3

    Science.gov (United States)

    Schmitt, Harrison H.

    1999-01-01

    Global demand for energy will likely increase by a factor of six or eight by the mid-point of the 21st Century due to a combination of population increase, new energy intensive technologies, and aspirations for improved standards of living in the less-developed world (1). Lunar helium-3 (3He), with a resource base in the Tranquillitatis titanium-rich lunar maria (2,3) of at least 10,000 tonnes (4), represents one potential energy source to meet this rapidly escalating demand. The energy equivalent value of 3He delivered to operating fusion power plants on Earth would be about 3 billion per tonne relative to today's coal which supplies most of the approximately 90 billion domestic electrical power market (5). These numbers illustrate the magnitude of the business opportunity. The results from the Lunar Prospector neutron spectrometer (6) suggests that 3He also may be concentrated at the lunar poles along with solar wind hydrogen (7). Mining, extraction, processing, and transportation of helium to Earth requires new innovations in engineering but no known new engineering concepts (1). By-products of lunar 3He extraction, largely hydrogen, oxygen, and water, have large potential markets in space and ultimately will add to the economic attractiveness of this business opportunity (5). Inertial electrostatic confinement (IEC) fusion technology appears to be the most attractive and least capital intensive approach to terrestrial fusion power plants (8). Heavy lift launch costs comprise the largest cost uncertainty facing initial business planning, however, many factors, particularly long term production contracts, promise to lower these costs into the range of 1-2000 per kilogram versus about 70,000 per kilogram fully burdened for the Apollo Saturn V rocket (1). A private enterprise approach to developing lunar 3He and terrestrial IEC fusion power would be the most expeditious means of realizing this unique opportunity (9). In spite of the large, long-term potential

  20. Relation between the conditions of helium ion implantation and helium void equilibrium parameters

    International Nuclear Information System (INIS)

    Neklyudov, I.M.; Rybalko, V.F.; Ruzhitskij, V.V.; Tolstolutskaya, G.D.

    1981-01-01

    The conditions of helium thermodynamic equilibrium in a system of voids produced by helium ion bombardment of a metal sample are studied. As an initial equation for description of the equilibrium the Clapeyron equation was used. The equation is obtained relating basic parameters of helium voids (average diameter and density) to irradiation parameters (dose, ion energy (straggling)) and properties of the metal (surface tension coefficient, yield strength). Comparison of the calculations with experimental data on helium in nickel found in literature shows that the equation yields satisfactory resutls for the dose range 1.10 16 -1x10 17 cm -2 and temperatures T [ru

  1. CALCULATED REGENERATOR PERFORMANCE AT 4 K WITH HELIUM-4 AND HELIUM-3

    International Nuclear Information System (INIS)

    Radebaugh, Ray; Huang Yonghua; O'Gallagher, Agnes; Gary, John

    2008-01-01

    The helium-4 working fluid in regenerative cryocoolers operating with the cold end near 4 K deviates considerably from an ideal gas. As a result, losses in the regenerator, given by the time-averaged enthalpy flux, are increased and are strong functions of the operating pressure and temperature. Helium-3, with its lower boiling point, behaves somewhat closer to an ideal gas in this low temperature range and can reduce the losses in 4 K regenerators. An analytical model is used to find the fluid properties that strongly influence the regenerator losses as well as the gross refrigeration power. The thermodynamic and transport properties of helium-3 were incorporated into the latest NIST regenerator numerical model, known as REGEN3.3, which was used to model regenerator performance with either helium-4 or helium-3. With this model we show how the use of helium-3 in place of helium-4 can improve the performance of 4 K regenerative cryocoolers. The effects of operating pressure, warm-end temperature, and frequency on regenerators with helium-4 and helium-3 are investigated and compared. The results are used to find optimum operating conditions. The frequency range investigated varies from 1 Hz to 30 Hz, with particular emphasis on higher frequencies

  2. Diffusion of helium and nucleation-growth of helium-bubbles in metallic materials

    International Nuclear Information System (INIS)

    Zhang Chonghong; Chen Keqin; Wang Yinshu

    2001-01-01

    Studies of diffusion and aggregation behaviour of helium in metallic materials are very important to solve the problem of helium embrittlement in structural materials used in the environment of nuclear power. Experimental studies on helium diffusion and aggregation in austenitic stainless steels in a wide temperature range have been performed in authors' research group and the main results obtained are briefly summarized. The mechanism of nucleation-growth of helium-bubbles has been discussed and some problems to be solved are also given

  3. Charged condensate and helium dwarf stars

    Energy Technology Data Exchange (ETDEWEB)

    Gabadadze, Gregory; Rosen, Rachel A, E-mail: gg32@nyu.edu, E-mail: rar339@nyu.edu [Center for Cosmology and Particle Physics, Department of Physics, New York University, New York, NY 10003 (United States)

    2008-10-15

    White dwarf stars composed of carbon, oxygen and heavier elements are expected to crystallize as they cool down below certain temperatures. Yet, simple arguments suggest that the helium white dwarf cores may not solidify, mostly because of zero-point oscillations of the helium ions that would dissolve the crystalline structure. We argue that the interior of the helium dwarfs may instead form a macroscopic quantum state in which the charged helium-4 nuclei are in a Bose-Einstein condensate, while the relativistic electrons form a neutralizing degenerate Fermi liquid. We discuss the electric charge screening, and the spectrum of this substance, showing that the bosonic long-wavelength fluctuations exhibit a mass gap. Hence, there is a suppression at low temperatures of the boson contribution to the specific heat-the latter being dominated by the specific heat of the electrons near the Fermi surface. This state of matter may have observational signatures.

  4. On the helium gas leak test

    International Nuclear Information System (INIS)

    Nishikawa, Akira; Ozaki, Susumu

    1975-01-01

    The helium gas leak test (Helium mass spectrometer testing) has a leak detection capacity of the highest level in practical leak tests and is going to be widely applied to high pressure vessels, atomic and vacuum equipments that require high tightness. To establish a standard test procedure several series of experiments were conducted and the results were investigated. The conclusions are summarized as follows: (1) The hood method is quantitatively the most reliable method. The leak rate obtained by tests using 100% helium concentration should be the basis of the other method of test. (2) The integrating method, bell jar method, and vacuum spray method can be considered quantitative when particular conditions are satisfied. (3) The sniffer method is not to be considered quantitive. (4) The leak rate of the hood, integrating, and bell jar methods is approximately proportional to the square of the helium partial pressure. (auth.)

  5. Helium-Hydrogen Recovery System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Immense quantities of expensive liquefied helium are required at Stennis and Kennedy Space Centers for pre-cooling rocket engine propellant systems prior to filling...

  6. KSTAR Helium Refrigeration System Design and Manufacturing

    International Nuclear Information System (INIS)

    Dauguet, P.; Briend, P.; Abe, I.; Fauve, E.; Bernhardt, J.-M.; Andrieu, F.; Beauvisage, J.

    2006-01-01

    The tokamak developed in the KSTAR (Korean Superconducting Tokamak Advanced Research) project makes intensive use of superconducting magnets operated at 4.5 K. The cold components of the KSTAR tokamak require forced flow of supercritical helium for magnets/structure, boiling liquid helium for current leads, and gaseous helium for thermal shields. The cryogenic system will provide stable operation and full automatic control. A three-pressure helium cycle composed of six turbines has been customised design for this project. The '' design '' operating mode results with a system composed of a 9 kW refrigerator (including safety margin) and using gas and liquid storages for mass balancing. During Shot/Standby mode, the heat loads are highly time-dependent. A thermal damper is used to smooth these variations and will allow stable operation. (author)

  7. Realization of mechanical rotation in superfluid helium

    Science.gov (United States)

    Gordon, E. B.; Kulish, M. I.; Karabulin, A. V.; Matyushenko, V. I.; Dyatlova, E. V.; Gordienko, A. S.; Stepanov, M. E.

    2017-09-01

    The possibility of using miniaturized low-power electric motors submerged in superfluid helium for organization of rotation inside a cryostat has been investigated. It has been revealed that many of commercial micromotors can operate in liquid helium consuming low power. Turret with 5 sample holders, assembled on the base of stepper motor, has been successfully tested in experiments on the nanowire production in quantized vortices of superfluid helium. Application of the stepper motor made it possible in a single experiment to study the effect of various experimental parameters on the yield and quality of the nanowires. The promises for continuous fast rotation of the bath filled by superfluid helium by using high-speed brushless micromotor were outlined and tested. Being realized, this approach will open new possibility to study the guest particles interaction with the array of parallel linear vortices in He II.

  8. Helium leak testing the Westinghouse LCP coil

    International Nuclear Information System (INIS)

    Merritt, P.A.; Attaar, M.H.; Hordubay, T.D.

    1983-01-01

    The tests, equipment, and techniques used to check the Westinghouse LCP coil for coolant flow path integrity and helium leakage are unique in terms of test sensitivity and application. This paper will discuss the various types of helium leak testing done on the LCP coil as it enters different stages of manufacture. The emphasis will be on the degree of test sensitivity achieved under shop conditions, and what equipment, techniques and tooling are required to achieve this sensitivity (5.9 x 10 -8 scc/sec). Other topics that will be discussed are helium flow and pressure drop testing which is used to detect any restrictions in the flow paths, and the LCP final acceptance test which is the final leak test performed on the coil prior to its being sent for testing. The overall allowable leak rate for this coil is 5 x 10 -6 scc/sec. A general evaluation of helium leak testing experience are included

  9. Near field characteristics of buoyant helium plumes

    Indian Academy of Sciences (India)

    pressure tubing. Helium gas enters the bottom of the settling chamber, passing through two ... A 40 mesh, flat stainless steel screen is placed across the orifice exit. ... PIV and PLIF measurements are carried out in phase resolved manner.

  10. Hybrid Circuit QED with Electrons on Helium

    Science.gov (United States)

    Yang, Ge

    Electrons on helium (eHe) is a 2-dimensional system that forms naturally at the interface between superfluid helium and vacuum. It has the highest measured electron mobility, and long predicted spin coherence time. In this talk, we will first review various quantum computer architecture proposals that take advantage of these exceptional properties. In particular, we describe how electrons on helium can be combined with superconducting microwave circuits to take advantage of the recent progress in the field of circuit quantum electrodynamics (cQED). We will then demonstrate how to reliably trap electrons on these devices hours at a time, at millikelvin temperatures inside a dilution refrigerator. The coupling between the electrons and the microwave resonator exceeds 1 MHz, and can be reproduced from the design geometry using our numerical simulation. Finally, we will present our progress on isolating individual electrons in such circuits, to build single-electron quantum dots with electrons on helium.

  11. Helium cosmic ray flux measurements at Mars

    International Nuclear Information System (INIS)

    Lee, Kerry; Pinsky, Lawrence; Andersen, Vic; Zeitlin, Cary; Cleghorn, Tim; Cucinotta, Frank; Saganti, Premkumar; Atwell, William; Turner, Ron

    2006-01-01

    The helium energy spectrum in Martian orbit has been observed by the MARIE charged particle spectrometer aboard the 2001 Mars Odyssey spacecraft. The orbital data were taken from March 13, 2002 to October 28, 2003, at which time a very intense Solar Particle Event caused a loss of communication between the instrument and the spacecraft. The silicon detector stack in MARIE is optimized for the detection of protons and helium in the energy range below 100MeV/n, which typically includes almost all of the flux during SPEs. This also makes MARIE an efficient detector for GCR helium in the energy range of 50-150MeV/n. We will present the first fully normalized flux results from MARIE, using helium ions in this energy range

  12. Helium cosmic ray flux measurements at Mars

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kerry [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States)]. E-mail: ktlee@ems.jsc.nasa.gov; Pinsky, Lawrence [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States); Andersen, Vic [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States); Zeitlin, Cary [National Space Biomedical Research Institute, Baylor College of Medicine, Houston, TX (United States); Cleghorn, Tim [NASA Johnson Space Center, 2101 NASA Road 1, Houston, TX 77058 (United States); Cucinotta, Frank [NASA Johnson Space Center, 2101 NASA Road 1, Houston, TX 77058 (United States); Saganti, Premkumar [Prairie View A and M University, P.O. Box 519, Prairie View, TX 77446-0519 (United States); Atwell, William [The Boeing Company, Houston, TX (United States); Turner, Ron [Advancing National Strategies and Enabling Results (ANSER), Arlington, Virginia (United States)

    2006-10-15

    The helium energy spectrum in Martian orbit has been observed by the MARIE charged particle spectrometer aboard the 2001 Mars Odyssey spacecraft. The orbital data were taken from March 13, 2002 to October 28, 2003, at which time a very intense Solar Particle Event caused a loss of communication between the instrument and the spacecraft. The silicon detector stack in MARIE is optimized for the detection of protons and helium in the energy range below 100MeV/n, which typically includes almost all of the flux during SPEs. This also makes MARIE an efficient detector for GCR helium in the energy range of 50-150MeV/n. We will present the first fully normalized flux results from MARIE, using helium ions in this energy range.

  13. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-01-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study

  14. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  15. Helium compressor aerodynamic design considerations for MHTGR circulators

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1988-01-01

    Compressor aerodynamic design considerations for both the main and shutdown cooling circulators in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) plant are addressed in this paper. A major selection topic relates to the impeller type (i.e., axial or radial flow), and the aerothermal studies leading to the selection of optimum parameters are discussed. For the conceptual designs of the main and shutdown cooling circulators, compressor blading geometries were established and helium gas flow paths defined. Both circulators are conservative by industrial standards in terms of aerodynamic and structural loading, and the blade tip speeds are particularly modest. Performance characteristics are presented, and the designs embody margin to ensure that pressure-rise growth potential can be accomodated should the circuit resistance possibly increase as the plant design advances. The axial flow impeller for the main circulator is very similar to the Fort St. Vrain (FSV) helium compressor which performs well. A significant technology base exists for the MHTGR plant circulators, and this is highlighted in the paper. (author). 15 refs, 16 figs, 12 tabs

  16. Helium behaviour in implanted boron carbide

    Directory of Open Access Journals (Sweden)

    Motte Vianney

    2015-01-01

    Full Text Available When boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing. The determination of the helium diffusion coefficient was carried out using the 3He(d,p4He nuclear reaction (NRA method. From the evolution of the width of implanted 3He helium profiles (fluence 1 × 1015/cm2, 3 MeV corresponding to a maximum helium concentration of about 1020/cm3 as a function of annealing temperatures, an Arrhenius diagram was plotted and an apparent diffusion coefficient was deduced (Ea = 0.52 ± 0.11 eV/atom. The dynamic of helium clusters was observed by transmission electron microscopy (TEM of samples implanted with 1.5 × 1016/cm2, 2.8 to 3 MeV 4He ions, leading to an implanted slab about 1 μm wide with a maximum helium concentration of about 1021/cm3. After annealing at 900 °C and 1100 °C, small (5–20 nm flat oriented bubbles appeared in the grain, then at the grain boundaries. At 1500 °C, due to long-range diffusion, intra-granular bubbles were no longer observed; helium segregates at the grain boundaries, either as bubbles or inducing grain boundaries opening.

  17. Effects of helium impurities on superalloys

    International Nuclear Information System (INIS)

    Selle, J.E.

    1977-07-01

    A review of the literature on the effects of helium impurities on superalloys at elevated temperatures was undertaken. The actual effects of these impurities vary depending on the alloy, composition of the gas atmosphere, and temperature. In general, exposure in helium produces significant but not catastrophic changes in the structure and properties of the alloys. The effects of these treatments on the structure, creep, fatigue, and mechanical properties of the various alloys are reviewed and discussed. Suggestions for future work are presented

  18. Electrons on the surface of liquid helium

    International Nuclear Information System (INIS)

    Lambert, D.K.

    1979-05-01

    Spectroscopic techniques were used to study transitions of electrons between bound states in the potential well near a helium surface. The charge density distribution of electrons on the surface was independently obtained from electrical measurements. From the measurements, information was obtained both about the interaction of the bound state electrons with the surface of liquid helium and about local disorder in the positions of electrons on the surface

  19. Effect of helium on void swelling in vanadium

    International Nuclear Information System (INIS)

    Brimhall, J.L.; Simonen, E.P.

    1975-01-01

    Little difference in void microstructural swelling of vanadium is observed when helium is injected simultaneously with a 46- or 5-MeV nickel beam as compared to no helium injection, at least at high dose rates. At lower dose rates, a strong helium effect is seen when the helium is injected prior to heavy ion bombardment. The effect of the helium is shown to be a strong function of the overall displacement damage rate

  20. Regulations and instructions for RA reactor operation

    International Nuclear Information System (INIS)

    1989-01-01

    This regulatory guide consists of following 4 chapters: Description of the RA reactor, organization scheme, regulations for performing experiments; Regulations for staff on duty; Instructions for operating the vacuum systems, heavy water and helium systems; and evacuation in case of accident [sr

  1. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  2. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  3. Sonic Helium Detectors in the Fermilab Tevatron

    Science.gov (United States)

    Bossert, R. J.

    2006-04-01

    In the Fermilab Tevatron cryogenic system there are many remotely located low-pressure plate relief valves that must vent large volumes of cold helium gas when magnet quenches occur. These valves can occasionally stick open or not reseat completely, resulting in a large helium loss. As such, the need exists for a detector to monitor the relief valve's discharge area for the presence of helium. Due to the quantity needed, cost is an important factor. A unit has been developed and built for this purpose that is quite inexpensive. Its operating principle is based on the speed of sound, where two closely matched tubes operate at their acoustic resonant frequency. When helium is introduced into one of these tubes, the resulting difference in acoustic time of flight is used to trigger an alarm. At present, there are 39 of these units installed and operating in the Tevatron. They have detected many minor and major helium leaks, and have also been found useful in detecting a rise in the helium background in the enclosed refrigerator buildings. This paper covers the construction, usage and operational experience gained with these units over the last several years.

  4. Sonic helium detectors in the Fermilab Tevatron

    International Nuclear Information System (INIS)

    Bossert, R.J.; Fermilab

    2006-01-01

    In the Fermilab Tevatron cryogenic system there are many remotely located low-pressure plate relief valves that must vent large volumes of cold helium gas when magnet quenches occur. These valves can occasionally stick open or not reseat completely, resulting in a large helium loss. As such, the need exists for a detector to monitor the relief valve's discharge area for the presence of helium. Due to the quantity needed, cost is an important factor. A unit has been developed and built for this purpose that is quite inexpensive. Its operating principle is based on the speed of sound, where two closely matched tubes operate at their acoustic resonant frequency. When helium is introduced into one of these tubes, the resulting difference in acoustic time of flight is used to trigger an alarm. At present, there are 39 of these units installed and operating in the Tevatron. They have detected many minor and major helium leaks, and have also been found useful in detecting a rise in the helium background in the enclosed refrigerator buildings. This paper covers the construction, usage and operational experience gained with these units over the last several years

  5. Separation of compressor oil from helium

    International Nuclear Information System (INIS)

    Strauss, R.; Perrotta, K.A.

    1982-01-01

    Compression of helium by an oil-sealed rorary screw compressor entrains as much as 4000 parts per million by weight of liquid and vapor oil impurities in the gas. The reduction below about 0.1 ppm for cryogenic applications is discussed. Oil seperation equipment designed for compressed air must be modified significantly to produce the desired results with helium. The main differences between air and helium filtration are described. A description of the coalescers is given with the continuous coalescing of liquid mist from air or other gas illustrated. Oil vapor in helium is discussed in terms of typical compressor oils, experimental procedure for measuring oil vapor concentration, measured volatile hydrocarbons in the lubricants, and calculated concentration of oil vapor in Helium. Liquid oil contamination in helium gas can be reduced well below 0.1 ppm by a properly designed multiple state coalescing filter system containing graded efficiency filter elements. The oil vapor problem is best attached by efficiently treating the oil to remove most of the colatiles before charging the compressor

  6. [Fluid dynamics of supercritical helium within internally cooled cabled superconductors

    International Nuclear Information System (INIS)

    Van Sciver, S.W.

    1995-01-01

    The Applied Superconductivity Center of the University of Wisconsin-Madison proposes to conduct research on low temperature helium fluid dynamics as it applies to the cooling of internally cooled cabled superconductors (ICCS). Such conductors are used in fusion reactor designs including most of the coils in ITER. The proposed work is primarily experimental involving measurements of transient and steady state pressure drop in a variety of conductor configurations. Both model and prototype conductors for actual magnet designs will be investigated. The primary goal will be to measure and model the friction factor for these complex geometries. In addition, an effort will be made to study transient processes such as heat transfer and fluid expulsion associated with quench conditions

  7. Corrosion tests of high temperature alloys in impure helium

    International Nuclear Information System (INIS)

    Berka, Jan; Kalivodova, Jana; Vilemova, Monika; Skoumalova, Zuzana; Brabec, Petr

    2014-01-01

    Czech research organizations take part several projects concerning technologies and materials for advanced gas cooled reactors, as an example international project ARCHER supported by EU within FP7, also several national projects supported by Technology Agency of the Czech Republic are solved in cooperation with industrial and research organization. Within these projects the material testing program is performed. The results presented in these paper concerning high temperature corrosion and degradation of alloys (800 H, SS 316 and P91) in helium containing minor impurities (H_2, CO, CH_4, HZO) at temperatures up to 760°C. After corrosion tests (up to 1500 hours) the specimens was investigated by several methods (gravimetry, SEM-EDX, optical microscopy, hardness and micro-hardness testing etc. (author)

  8. Martensitic transformation in helium implanted 316 stainless steel

    International Nuclear Information System (INIS)

    Ishimatsu, Manabu; Tsukuda, Noboru

    1997-01-01

    In order to simulate surface deterioration phenomenon due to particle loading of SUS-316 steel which is one of candidate materials for nuclear fusion reactor vacuum wall structure material, helium ion implanting was conducted at room temperature, 473 K and 573 K. To martensitic phase formed as a results, implantation dose dependence, implanting temperature dependence, and annealing under 1073 K were conducted. Formation of the martensitic phase was suppressed at high implanting temperature. At room temperature implantation, the martensitic phase disappeared at more than 873 K, but at high temperature implantation, it increased abnormally near at 973 K. This showed that deterioration of materials depended extremely upon using temperature and temperature history. (G.K.)

  9. Experimental investigations of helium cryotrapping by argon frost

    International Nuclear Information System (INIS)

    Mack, A.; Perinic, D.; Murdoch, D.; Boissin, J.C.

    1992-01-01

    At the Karlsruhe Nuclear Research Centre (KfK) cryopumping techniques are being investigated by which the gaseous exhausts from the NET/ITER reactor can be pumped out during the burn-and dwell-times. Cryosorption and cryotrapping are techniques which are suitable for this task. It is the target of the investigations to test the techniques under NET/ITER conditions and to determine optimum design data for a prototype. They involve measurement of the pumping speed as a function of the gas composition, gas flow and loading condition of the pump surfaces. The following parameters are subjected to variations: Ar/He ratio, specific helium volume flow rate, cryosurface temperature, process gas composition, impurities in argon trapping gas, three-stage operation and two-stage operation. This paper is a description of the experiments on argon trapping techniques started in 1990. Eleven tests as well as the results derived from them are described

  10. Effects of helium implantation on fatigue properties of F82H-IEA heat

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, N.; Murase, Y.; Nagakawa, J. [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)

    2007-07-01

    Full text of publication follows: Ferritic steels including reduced activation ones that have been recognized as attractive structural candidates for DEMO reactors and the beyond are known to be highly resistant to helium embrittlement. However, almost studies that deduced this behavior have been carried out by means of short time experiments such as tensile tests, and a few results are available concerning long term inspections, although the detrimental helium effect appears more severely in the latter. The aim of this work is to obtain further information on the influence of helium on fatigue properties of a representative reduced activation ferritic/martensitic steel F82H (8Cr2WVTa) using helium implantation technique with a cyclotron. The material examined is an IEA heat version of F82H. In order to realize a fine grain size due to thin specimens (0.08 mm thick) for ion irradiation, normalizing was conducted at rather low temperature of 1213 K, followed by tempering at 1023 K. Helium was implanted by {alpha}-particle irradiation at 823 K, a desired highest temperature of this material for first wall application, to the concentration of 100 appm He with an implantation rate of about 1.7 x 10{sup -3} appm He/s. Subsequent fatigue tests were conducted at the same temperature as that of irradiation, not only on implanted specimens but also on reference controls which were not implanted with helium but experienced the same metallurgical histories as those of irradiated ones. After fracture, samples were observed with electron microscopes. In short time periods, it has been notified that helium introduction caused no significant deterioration of both fatigue life and extension at fracture. In addition, all specimens failed in a fully trans-crystalline and ductile manner, irrespective of whether helium was present or not. Indication of grain boundary embrittlement was therefore not discerned. These facts would reflect insusceptible characteristics of this material to

  11. Effects of helium implantation on fatigue properties of F82H-IEA heat

    International Nuclear Information System (INIS)

    Yamamoto, N.; Murase, Y.; Nagakawa, J.

    2007-01-01

    Full text of publication follows: Ferritic steels including reduced activation ones that have been recognized as attractive structural candidates for DEMO reactors and the beyond are known to be highly resistant to helium embrittlement. However, almost studies that deduced this behavior have been carried out by means of short time experiments such as tensile tests, and a few results are available concerning long term inspections, although the detrimental helium effect appears more severely in the latter. The aim of this work is to obtain further information on the influence of helium on fatigue properties of a representative reduced activation ferritic/martensitic steel F82H (8Cr2WVTa) using helium implantation technique with a cyclotron. The material examined is an IEA heat version of F82H. In order to realize a fine grain size due to thin specimens (0.08 mm thick) for ion irradiation, normalizing was conducted at rather low temperature of 1213 K, followed by tempering at 1023 K. Helium was implanted by α-particle irradiation at 823 K, a desired highest temperature of this material for first wall application, to the concentration of 100 appm He with an implantation rate of about 1.7 x 10 -3 appm He/s. Subsequent fatigue tests were conducted at the same temperature as that of irradiation, not only on implanted specimens but also on reference controls which were not implanted with helium but experienced the same metallurgical histories as those of irradiated ones. After fracture, samples were observed with electron microscopes. In short time periods, it has been notified that helium introduction caused no significant deterioration of both fatigue life and extension at fracture. In addition, all specimens failed in a fully trans-crystalline and ductile manner, irrespective of whether helium was present or not. Indication of grain boundary embrittlement was therefore not discerned. These facts would reflect insusceptible characteristics of this material to high

  12. Recombination of positive helium ions in gaseous helium

    International Nuclear Information System (INIS)

    Shyu, J.S.

    1988-01-01

    The Wigner-Keck Monte Carlo trajectory method and the resonance complex theory are employed to calculate the rate coefficient k for H e + ions recombining in gaseous helium in the temperature range 80 2 + is obtained from a Morse potential and a long range ion-induced dipole interaction term. The three body He 3 + interaction is represented by an approximate expression which, for practical purpose, depends on the same parameters that determine the two body interaction. Russell had employed the Wigner-Keck Monte Carlo trajectory method to the same reaction. Unlike his calculation, in which the final quasibound states are treated as continuous, we apply the JWKB approximation to quantize those quasibound states. Both the values of k, calculated from two different quasibound state treatments, are found to be very close and give good agreement with experimental results obtained by Biondi, although they are still 10% to 20% lower than the experimental results. The resonance complex theory, developed by Roberts et al, is then employed to investigated de-excitation from the highest quasibound state, which can be populated by inward tunneling through the rotational (centrifugal) barrier. It is found that this strongly supports a suggestion proposed by Russell. He had suggested that the remaining difference between the Wigner-Keck method and experiment might be largely due to the formation of highly excited quasibound states. The statistical errors of the rate constants, which is the sun of results obtained from both methods, are kept less then 5% by running 2500 trajectories in the first method and 500 in the second

  13. Helium mobility in advanced nuclear ceramics

    International Nuclear Information System (INIS)

    Agarwal, Shradha

    2014-01-01

    The main goal of this work is to improve our knowledge on the mechanisms able to drive the helium behaviour in transition metal carbides and nitrides submitted to thermal annealing or ion irradiation. TiC, TiN and ZrC polycrystals were implanted with 3 MeV 3 He ions at room temperature in the fluence range 2 * 10 15 et 6 * 10 16 cm -2 . Some of them have been pre-irradiated with self-ions (14 MeV Ti or Zr). Fully controlled thermal annealing tests were subsequently carried out in the temperature range 1000 - 1600 C for two hours. The evolution of the helium depth distribution in function of implantation dose, temperature and pre-irradiation dose was measured thanks to the deuteron-induced nuclear reaction 3 He(d, p 0 ) 4 He between 900 keV and 1.8 MeV. The microstructure of implanted and annealed samples was investigated by transmission electron microscopy on thin foils prepared using the FIB technique. Additional characterization tools, as X-ray diffraction and Raman microspectrometry, have been also applied in order to obtain complementary information. Among the most relevant results obtained, the following have to be outlined: - double-peak helium depth profile was measured on as implanted sample for the three compounds. The first peak is located near the end of range and includes the major part of helium, a second peak located close to the surface corresponds to the helium atoms trapped by the native vacancies; - the helium retention capacity in transition metal carbides and nitrides submitted to fully controlled thermal treatments varies according to ZrC 0.92 ≤ TiC 0.96 ≤ TiN 0.96 ; - whatever the investigated material, a self-ion-induced pre-damaging does not modify the initial helium profile extent. The influence of the post-implantation thermal treatment remains preponderant in any case; - the apparent diffusion coefficient of helium is in the range 4 * 10 -18 - 2 * 10 -17 m 2 s -1 in TiC0.96 and 3.5 * 10 -19 - 5.3 * 10 -18 m 2 s -1 in TiN 0.96 between

  14. Experimental mechanistic investigation of the nanostructuring of tungsten with low energy helium plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Fiflis, P., E-mail: fiflis1@illinois.edu; Connolly, N.; Ruzic, D.N.

    2016-12-15

    Helium ion bombardment of tungsten at temperatures between approximately one third and one half of its melting point has shown growth of nanostructures colloquially referred to as “fuzz”. The nanostructures take the form of thin tendrils of diameter about 30 nm and grow out of the bulk material. Tungsten will and does compose one of the key materials for plasma facing components (PFCs) in fusion reactors. The formation of nanostructured fuzz layers on PFCs would be detrimental to the performance of the reactor, and must therefore be avoided. Previous experiments have shown evidence that tungsten fuzz is initially grown by loop punching of helium bubbles created in the bulk. However, once the tendrils grow to sufficient length, the tendrils should intercept the entire helium flux, halting the production of fuzz. Fuzz continues to grow though. To increase the understanding of the mechanisms of tungsten fuzz formation, and thereby aid the avoidance of its production, a series of tests were performed to examine the validity of several theories regarding later stage tungsten fuzz growth. Tests showed that the fuzz formation was dependent solely on the bombardment of helium ions, and not on electric fields, or adatom diffusion. Experiments employing a tungsten coated molybdenum sample indicate the presence of a strong mixing layer and strongly suggest that tungsten fuzz growth continues to occur from the bottom up even as the tendrils grow in size. Tests also show a similarity between different metals exposed to helium ion fluxes where the ratio of bubble diameter to tendril diameter is constant.

  15. Experimental mechanistic investigation of the nanostructuring of tungsten with low energy helium plasmas

    International Nuclear Information System (INIS)

    Fiflis, P.; Connolly, N.; Ruzic, D.N.

    2016-01-01

    Helium ion bombardment of tungsten at temperatures between approximately one third and one half of its melting point has shown growth of nanostructures colloquially referred to as “fuzz”. The nanostructures take the form of thin tendrils of diameter about 30 nm and grow out of the bulk material. Tungsten will and does compose one of the key materials for plasma facing components (PFCs) in fusion reactors. The formation of nanostructured fuzz layers on PFCs would be detrimental to the performance of the reactor, and must therefore be avoided. Previous experiments have shown evidence that tungsten fuzz is initially grown by loop punching of helium bubbles created in the bulk. However, once the tendrils grow to sufficient length, the tendrils should intercept the entire helium flux, halting the production of fuzz. Fuzz continues to grow though. To increase the understanding of the mechanisms of tungsten fuzz formation, and thereby aid the avoidance of its production, a series of tests were performed to examine the validity of several theories regarding later stage tungsten fuzz growth. Tests showed that the fuzz formation was dependent solely on the bombardment of helium ions, and not on electric fields, or adatom diffusion. Experiments employing a tungsten coated molybdenum sample indicate the presence of a strong mixing layer and strongly suggest that tungsten fuzz growth continues to occur from the bottom up even as the tendrils grow in size. Tests also show a similarity between different metals exposed to helium ion fluxes where the ratio of bubble diameter to tendril diameter is constant.

  16. Experimental and numerical thermohydraulic study of a supercritical helium loop in forced convection under pulsed heat loads

    International Nuclear Information System (INIS)

    Lagier, Benjamin

    2014-01-01

    Future fusion reactor devices such as ITER or JT-60SA will produce thermonuclear fusion reaction in plasmas at several millions of degrees. The confinement in the center of the chamber is achieved by very intense magnetic fields generated by superconducting magnets. These coils have to be cooled down to 4.4 K through a forced flow of supercritical helium. The cyclic behavior of the machines leads to pulsed thermal heat loads which will have to be handled by the refrigerator. The HELIOS experiment built in CEA Grenoble is a scaled down model of the helium distribution system of the tokamak JT-60SA composed of a saturated helium bath and a supercritical helium loop. The thesis work explores HELIOS capabilities for experimental and numerical investigations on three heat load smoothing strategies: the use of the saturated helium bath as an open thermal buffer, the rotation speed variation of the cold circulator and the bypassing of the heated section. The developed model describes well the physical evolutions of the helium loop (pressure, temperature, mass flow) submitted to heat loads observed during experiments. Advanced controls have been tested and validated to improve the stability of the refrigerator and to optimize the refrigeration power. (author) [fr

  17. Helium gas turbine conceptual design by genetic/gradient optimization

    International Nuclear Information System (INIS)

    Yang, Long; Yu, Suyuan

    2003-01-01

    Helium gas turbine is the key component of the power conversion system for direct cycle High Temperature Gas-cooled Reactors (HTGR), of which an optimal design is essential for high efficiency. Gas turbine design currently is a multidisciplinary process in which the relationships between constraints, objective functions and variables are very noisy. Due to the ever-increasing complexity of the process, it has becomes very hard for the engineering designer to foresee the consequences of changing certain parts. With classic design procedures which depend on adaptation to baseline design, this problem is usually averted by choosing a large number of design variables based on the engineer's judgment or experience in advance, then reaching a solution through iterative computation and modification. This, in fact, leads to a reduction of the degree of freedom of the design problem, and therefore to a suboptimal design. Furthermore, helium is very different in thermal properties from normal gases; it is uncertain whether the operation experiences of a normal gas turbine could be used in the conceptual design of a helium gas turbine. Therefore, it is difficult to produce an optimal design with the general method of adaptation to baseline. Since their appearance in the 1970s, Genetic algorithms (GAs) have been broadly used in many research fields due to their robustness. GAs have also been used recently in the design and optimization of turbo-machines. Researchers at the General Electronic Company (GE) developed an optimization software called Engineous, and used GAs in the basic design and optimization of turbines. The ITOP study group from Xi'an Transportation University also did some work on optimization of transonic turbine blades. However, since GAs do not have a rigorous theory base, many problems in utilities have arisen, such as premature convergence and uncertainty; the GA doesn't know how to locate the optimal design, and doesn't even know if the optimal solution

  18. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  19. 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.

  20. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

    The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

  1. Towards helium-3 neutron polarizers

    International Nuclear Information System (INIS)

    Tasset, F.

    1995-01-01

    With a large absorption cross-section entirely due to antiparallel spin capture, polarized helium-3 is presently the most promising broad-band polarizer for thermal and epithermal neutrons. Immediate interest was raised amongst the neutron community when a dense gaseous 3 He polarizer was used for the first time in 1988, on a pulsed neutron beam at Los Alamos. With 20 W of laser power on a 30 cm long, 8.6 atm target, 40% 3 He polarization was achieved in a recent polarized electron scattering experiment at SLAC. In this technique the 3 He nuclei are polarized directly at an appropriate high pressure through spin-exchange collisions with a thick, optically pumped rubidium vapor. A different and competitive approach is being presently developed at Mainz University in collaboration with ENS Paris and now the ILL. A discharge is established in pure 3 He at low pressure producing excited metastable atoms which can be optically pumped with infra-red light. Highly effective exchange collision with the atoms remaining in the ground state quickly produces 75% polarization at 1.5 mbar. A truly non-magnetic system then compresses the polarized gas up to several bars as required. The most recent machine comprises a two-stage glass-titanium compressor. In less than 1 h it can inflate a 100 cm 3 target cell with three bars of polarized gas. The very long relaxation times (several days) now being obtained at high pressure with a special metallic coating on the glass walls, the polarized cell can be detached and inserted in the neutron beam as polarizer. We expect 50% 3 He-polarization to be reached soon, allowing such filters to compete favorably with existing Heusler-crystal polarizers at thermal and short neutron wavelengths. It must be stressed that such a system based on a 3 He polarization factory able to feed several passive, transportable, polarizers is well matched to neutron scattering needs. (orig.)

  2. Rotons, Superfluidity, and Helium Crystals

    Science.gov (United States)

    Balibar, Sébastien

    2006-09-01

    Fritz London understood that quantum mechanics could show up at the macroscopic level, and, in 1938, he proposed that superfluidity was a consequence of Bose-Einstein condensation. However, Lev Landau never believed in London's ideas; instead, he introduced quasiparticles to explain the thermodynamics of superfluid 4He and a possible mechanism for its critical velocity. One of these quasiparticles, a crucial one, was his famous "roton" which he considered as an elementary vortex. At the LT0 conference (Cambridge, 1946), London criticized Landau and his "theory based on the shaky grounds of imaginary rotons". Despite their rather strong disagreement, Landau was awarded the London prize in 1960, six years after London's death. Today, we know that London and Landau had both found part of the truth: BEC takes place in 4He, and rotons exist. In my early experiments on quantum evaporation, I found direct evidence for the existence of rotons and for evaporation processes in which they play the role of photons in the photoelectric effect. But rotons are now considered as particular phonons which are nearly soft, due to some local order in superfluid 4He. Later we studied helium crystals which are model systems for the general study of crystal surfaces, but also exceptional systems with unique quantum properties. In our recent studies of nucleation, rotons show their importance again: by using acoustic techniques, we have extended the study of liquid 4He up to very high pressures where the liquid state is metastable, and we wish to demonstrate that the vanishing of the roton gap may destroy superfluidity and trigger an instability towards the crystalline state.

  3. High-temperature helium-loop facility

    International Nuclear Information System (INIS)

    Tokarz, R.D.

    1981-09-01

    The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

  4. Microstructural observation on helium injected and creep ruptured JPCA

    International Nuclear Information System (INIS)

    Yamamoto, N.; Shiraishi, H.; Hishinuma, A.

    1986-01-01

    Detailed and quantitative TEM observation was performed on high temperature helium injected and creep ruptured JPCA to seek the prominent TiC distribution developed for suppression of helium embrittlement. Three different preinjection treatments were adopted for changing the TiC distribution. Considerable degradation in creep rupture strength by helium occurred in solution-annealed specimens, although there was much less effect of other treatments which included aging prior to injection. The concentration of helium at grain boundaries and the promotion of precipitation by helium during injection were responsible for the degradation. Therefore, the presence of TiC precipitates before helium introduction will help prevent degradation. On the other hand, the rupture elongation was reduced by helium after all treatments, although helium trapping by TiC precipitates in the matrix was successfully achieved. Consequently, the combined use of several methods may be necessary for further suppression of helium embrittlement. (orig.)

  5. Resistivity studies of interstitial helium mobility in niobium

    International Nuclear Information System (INIS)

    Chen, C.G.; Birnbaum, H.K.; Johnson, A.B. Jr.

    1979-01-01

    The mobility of interstitial helium in Nb and Nb-O alloys was studied in the temperature range of 10-383 K using resistivity measurements. The helium was introduced by radioactive decay of solute tritium (approximately 1 at%). At T < 100 K the resistivity increased due to conversion of tritium trapped at oxygen interstititals to helium. The formation of helium caused a very significant resistance increase at room temperature and above. The results suggest that helium is mobile at temperatures above 295 K and that the precipitation of large helium bubbles occurs along grain boundaries. The mobile helium species may either be single interstitials or small helium clusters. The activation enthalpy for the diffusion of the mobile helium species was estimated to be about 55 kJ/mol (0.66 eV). (Auth.)

  6. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

    1997-01-01

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  7. Quantum statistics and liquid helium 3 - helum 4 mixtures

    International Nuclear Information System (INIS)

    Cohen, E.G.D.

    1979-01-01

    The behaviour of liquid helium 3-helium 4 mixtures is considered from the point of view of manifestation of quantum statistics effects in macrophysics. The Boze=Einstein statistics is shown to be of great importance for understanding superfluid helium-4 properties whereas the Fermi-Dirac statistics is of importance for understanding helium-3 properties. Without taking into consideration the interaction between the helium atoms it is impossible to understand the basic properties of liquid helium 33 - helium 4 mixtures at constant pressure. Proposed is a simple model of the liquid helium 3-helium 4 mixture, namely the binary mixture consisting of solid spheres of two types subjecting to the Fermi-Dirac and Bose-Einstein statistics relatively. This model predicts correctly the most surprising peculiarities of phase diagrams of concentration dependence on temperature for helium solutions. In particular, the helium 4 Bose-Einstein statistics is responsible for the phase lamination of helium solutions at low temperatures. It starts in the peculiar critical point. The helium 4 Fermi-Dirac statistics results in incomplete phase lamination close to the absolute zero temperatures, that permits operation of a powerful cooling facility, namely refrigerating machine on helium solution

  8. Cover gases in nuclear reactors with special reference to argon

    International Nuclear Information System (INIS)

    Jose, C.J.; Shah, G.C.; Prabhu, L.H.; Vartak, D.G.

    1975-01-01

    The report describes the specifications of an ideal cover gas for the smooth operation of a nuclear reactor. The advantages of using helium as cover gas, the sources of impurities in helium cover gas and the methods of purification of helium are given in detail. Various problems associated with the use of argon as cover gas and methods to purify and decontaminate argon cover gas are discussed on the basis of experimental data collected. A laboratory model of the system which can be used to evaluate the performance of the gas purification adsorbents is also described. (author)

  9. In pile helium loop ''Comedie''

    International Nuclear Information System (INIS)

    Blanchard, R.J.

    1985-01-01

    The loop is located in the SILOE reactor at Centre d'Etudes Nucleaires de Grenoble. The purpose and objectives are divided into two groups, principal and secondary. The primary objective was to provide basic data on the deposition behavior of important condensable fission products on a variety of steel surfaces, i.e. temperature (sorption isotherms) and mass transfer (physical adsorption) dependencies; to provide information concerning the degree of penetration of important fission products into the metals comprising the heat exchanger-recuperator tubes as a function of alloy type and/or metal temperature; to provide complementary information on the reentrainment (liftoff) of important fission and activation products by performing out-of-pile blowdown experiments on tube samples representative of the alloy types used in the heat exchanger-recuperator and of the surface temperatures experienced during plateout. The secondary objective was to provide information concerning the migration of important fission products through graphite. To this end, concentration profiles in the web between the fuel rods containing the fission product source and the coolant channels and in the graphite diffusion sample will be measured to study the corrosion of metallic specimens placed in the conditions of high temperature gas cooled reactor. The first experiment SRO enables to determine the loop characteristics and possibilities related to thermal, thermodynamic, chemical and neutronic properties. The second experiment has been carried out in high temperature gas cooled reactor operating conditions. It enables to determine in particular the deposition axial profile of activation and fission products in the plateout section constituting the heat exchanger, the fission products balance trapped in the different filter components, and the cumulated released fraction of solid fission products. The SR1 test permits to demonstrate in particular the Comedie loop operation reliability, either

  10. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    Hogan, W.J.; Pitts, J.H.

    1986-01-01

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO 2 . The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  11. Nuclear reactor plant with a small gas-cooled HT reactor accommodated in a steel pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.

    1986-01-01

    The plant has a small HT reactor and an He/He heat exchanger situated above this, with preferably two parallel circulating blowers connected after it. It also has at least one post-shutdown heat removal system, which is situated after the He/He heat exchanger in the direction of flow and which always has the total quantity of primary helium flowing through it. In one version of the design, the heat exchanger consists of two concentric bundles of helices connected after one another, which have primary helium flowing in one direction and secondary helium in the opposite direction. (orig./HP) [de

  12. Helium release from radioisotope heat sources

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

    1984-05-01

    Diffusion of helium in /sup 238/PuO/sub 2/ fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel.

  13. Recent developments in liquid helium 3

    International Nuclear Information System (INIS)

    Ramarao, I.

    1977-01-01

    The current status of the theories for the ground state of liquid helium 3, are reviewed. To begin with, a brief summary of the experimental results on the thermodynamic properties of liquid helium 3 including its recently discovered superfulid phases is given. The basic ideas of the Landau theory of a normal Fermi liquid are then introduced. A qualitative discussion of the current understanding of the anisotropic phases of superfluid helium 3 is given, the microscopic calculaations for the binding energy of liquid helium 3 are reviewed and the results obtained for the two-body contributions to the binding energy using the Brueckner-Goldstone formulation and that of Mohling and his collaborators are summarized and discussed. The importance of a proper estimate of the three-body contributions to the binding energy is stressed. The results obtained in the literature using variational methods and constrained variational methods are discussed. A critical analysis of the results by various methods is given. Despite much effort the basic problem of the ground state of liquid helium 3, remains unresolved. (author)

  14. Film boiling heat transfer in liquid helium

    International Nuclear Information System (INIS)

    Inai, Nobuhiko

    1979-01-01

    The experimental data on the film boiling heat transfer in liquid helium are required for investigating the stability of superconducting wires. On the other hand, liquid helium has the extremely different physical properties as compared with the liquids at normal temperature such as water. In this study, the experiments on pool boiling were carried out, using the horizontal top surface of a 20 mm diameter copper cylinder in liquid helium. For observing individual bubbles, the experiments on film boiling from a horizontal platinum wire were performed separately in liquid nitrogen and liquid helium, and photographs of floating-away bubbles were taken. The author pointed out the considerable upward shift of the boiling curve near the least heat flux point in film boiling from the one given by the Berenson's equation which has been said to agree comparatively well with the data on the film boiling of the liquids at normal temperature, and the reason was investigated. Consequently, a model for film boiling heat transfer was presented. Also one equation expressing the film boiling at low heat flux for low temperature liquids was proposed. It represents well the tendency to shift from Berenson's equation of the experimental data on film boiling at the least heat flux point for liquid helium, liquid nitrogen and water having extremely different physical properties. Some discussions are added at the end of the paper. (Wakatsuki, Y.)

  15. Helium release from radioisotope heat sources

    International Nuclear Information System (INIS)

    Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

    1984-05-01

    Diffusion of helium in 238 PuO 2 fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel

  16. Helium, iron and electron particle transport and energy transport studies on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Grek, B.; Hill, K.W.; Hulse, R.A.; Johnson, D.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Redi, M.H.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor

  17. Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

    Science.gov (United States)

    Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

  18. Particle and energy transport studies on TFTR and implications for helium ash in future fusion devices

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Bell, R.E.; Grek, B.; Hulse, R.A.; Johnson, D.W.; Hill, K.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1992-01-01

    Particle and energy transport in tokamak plasmas have long been subjects of vigorous investigation. Present-day measurement techniques permit radially resolved studies of the transport of electron perturbations, low- and high-Z impurities, and energy. In addition, developments in transport theory provide tools that can be brought to bear on transport issues. Here, we examine local particle transport measurements of electrons, fully-stripped thermal helium, and helium-like iron in balanced-injection L-mode and enhanced confinement deuterium plasmas on TFTR of the same plasma current, toroidal field, and auxiliary heating power. He 2+ and Fe 24+ transport has been studied with charge exchange recombination spectroscopy, while electron transport has been studied by analyzing the perturbed electron flux following the same helium puff used for the He 2+ studies. By examining the electron and He 2+ responses following the same gas puff in the same plasmas, an unambiguous comparison of the transport of the two species has been made. The local energy transport has been examined with power balance analysis, allowing for comparisons to the local thermal fluxes. Some particle and energy transport results from the Supershot have been compared to a transport model based on a quasilinear picture of electrostatic toroidal drift-type microinstabilities. Finally, implications for future fusion reactors of the observed correlation between thermal transport and helium particle transport is discussed

  19. Analysis of helium purification system capability during water ingress accident in RDE

    Science.gov (United States)

    Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

  20. Graphite moderated reactor for thermoelectric generation

    International Nuclear Information System (INIS)

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

    Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

  1. Feasibility of lunar Helium-3 mining

    Science.gov (United States)

    Kleinschneider, Andreas; Van Overstraeten, Dmitry; Van der Reijnst, Roy; Van Hoorn, Niels; Lamers, Marvin; Hubert, Laurent; Dijk, Bert; Blangé, Joey; Hogeveen, Joel; De Boer, Lennaert; Noomen, Ron

    With fossil fuels running out and global energy demand increasing, the need for alternative energy sources is apparent. Nuclear fusion using Helium-3 may be a solution. Helium-3 is a rare isotope on Earth, but it is abundant on the Moon. Throughout the space community lunar Helium-3 is often cited as a major reason to return to the Moon. Despite the potential of lunar Helium-3 mining, little research has been conducted on a full end-to-end mission. This abstract presents the results of a feasibility study conducted by students from Delft University of Technology. The goal of the study was to assess whether a continuous end-to-end mission to mine Helium-3 on the Moon and return it to Earth is a viable option for the future energy market. The set requirements for the representative end-to-end mission were to provide 10% of the global energy demand in the year 2040. The mission elements have been selected with multiple trade-offs among both conservative and novel concepts. A mission architecture with multiple decoupled elements for each transportation segment (LEO, transfer, lunar surface) was found to be the best option. It was found that the most critical element is the lunar mining operation itself. To supply 10% of the global energy demand in 2040, 200 tons of Helium-3 would be required per year. The resulting regolith mining rate would be 630 tons per second, based on an optimistic concentration of 20 ppb Helium-3 in lunar regolith. Between 1,700 to 2,000 Helium-3 mining vehicles would be required, if using University of Wisconsin’s Mark III miner. The required heating power, if mining both day and night, would add up to 39 GW. The resulting power system mass for the lunar operations would be in the order of 60,000 to 200,000 tons. A fleet of three lunar ascent/descent vehicles and 22 continuous-thrust vehicles for orbit transfer would be required. The costs of the mission elements have been spread out over expected lifetimes. The resulting profits from Helium

  2. Investigation of helium-induced embrittlement

    International Nuclear Information System (INIS)

    Sabelova, V.; Slugen, V.; Krsjak, V.

    2014-01-01

    In this work, the hardness of Fe-9%(wt.) Cr binary alloy implanted by helium ions up to 1000 nm was investigated. The implantations were performed using linear accelerator at temperatures below 80 grad C. Isochronal annealing up to 700 grad C with the step of 100 grad C was applied on the helium implanted samples in order to investigate helium induced embrittlement of material. Obtained results were compared with theoretical calculations of dpa profiles. Due to the results, the nano-hardness technique results to be an appropriate approach to the hardness determination of thin layers of implanted alloys. Both, experimental and theoretical calculation techniques (SRIM) show significant correlation of measured results of induced defects. (authors)

  3. Fast reactor fluence dosimetry. Technical progress report, January--November 1976

    International Nuclear Information System (INIS)

    1976-01-01

    The objectives of this task are to: (1) develop and demonstrate the use of 10 B and 6 Li helium accumulation fluence monitors (HAFM's) as a reliable and accurate method of measuring reactor neutron fluence; (2) develop and apply an expanded set of HAFM's which will provide fluence responses in different but overlapping neutron energy ranges; (3) identify, through the precise measurement of spectrum-integrated helium production cross sections, those elements which produce significant helium when used individually or as components of advanced alloys in FTR and LMFBR neutron environments, so that their use might be eliminated, minimized, or controlled; (4) use this information to predict, with confidence, the helium production rate for any alloy or material considered for fast reactor use, and (5) maintain a centralized helium measurements laboratory available to the research community, and upgrade the sample throughput capacity to handle FTR dosimetry requirements

  4. Nucleation path of helium bubbles in metals during irradiation

    International Nuclear Information System (INIS)

    Morishita, Kazunori

    2008-01-01

    Thermodynamical formalization is made for description of the nucleation and growth of helium bubbles in metals during irradiation. The proposed formalization is available or evaluating both microstructural changes in fusion first wall materials where helium is produced by (n, α) nuclear transmutation reactions, and those in fusion divertor materials where helium particles with low energy are directly implanted. Calculated nucleation barrier is significantly reduced by the presence of helium, showing that a helium bubble with an appropriate number of helium atoms depending on bubble size can nucleate without any large nucleation barriers, even at a condition where an empty void has very large nucleation barrier without helium. With the proposed thermodynamical formalization, the nucleation and growth process of helium bubbles in iron during irradiation is simulated by the kinetic Monte-Carlo (KMC) technique. It shows the nucleation path of a helium bubble on the (N He , N V ) space as functions of temperatures and the concentration of helium in the matrix, where N He and N V are the number of helium atoms and vacancies in the helium bubble, respectively. Bubble growth rates depend on the nucleation path and suggest that two different mechanisms operate for bubble growth: one is controlled by vacancy diffusion and the other is controlled by interstitial helium diffusion. (author)

  5. Experience with reactor assembly of FBTR

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

    2006-01-01

    Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

  6. Adsorption pump for helium pumping out

    International Nuclear Information System (INIS)

    Donde, A.L.; Semenenko, Yu.E.

    1981-01-01

    Adsorption pump with adsorbent cooling by liquid helium is described. Shuttered shield protecting adsorbent against radiation is cooled with evaporating helium passing along the coil positioned on the shield. The pump is also equipped with primed cylindrical shield, cooled with liquid nitrogen. The nitrogen shield has in the lower part the shuttered shield, on the pump casing there is a valve used for pump pre-burning, and valves for connection to recipient as well. Pumping- out rates are presented at different pressures and temperatures of adsorbent. The pumping-out rate according to air at absorbent cooling with liquid nitrogen constituted 5x10 -4 Pa-3000 l/s, at 2x10 -2 Pa-630 l/s. During the absorbent cooling with liquid hydrogen the pumping-out rate according to air was at 4x10 -4 Pa-580 l/s, at 2x10 -3 Pa-680 l/s, according to hydrogen - at 8x10 -5 Pa-2500 l/s, at 5x10 -3 Pa-4200 l/s. During adsorbent cooling with liquid helium the rate of pumping-out according to hydrogen at 3x10 5 Pa-2400% l/s, at 6x10 3 Pa-1200 l/s, and according to helium at 3.5x10 -5 Pa-2800 l/s, at 4x10 -3 Pa-1150 l/s. The limit vacuum is equal to 1x10 -7 Pa. The volume of the vessel with liquid helium is equal to 3.5 l. Helium consumption is 80 cm 3 /h. Consumption of liquid nitrogen from the shield is 400 cm 3 /h. The limit pressure in the pump is obtained after forevacuum pumping-out (adsorbent regeneration) at 300 K temperature. The pump is made of copper. The pump height together with primed tubes is 800 mm diameter-380 mm [ru

  7. Modeling of helium effects in metals: High temperature embrittlement

    International Nuclear Information System (INIS)

    Trinkaus, H.

    1985-01-01

    The effects of helium on swelling, creep rupture and fatigue properties of fusion reactor materials subjected to (n,α)-reactions and/or direct α-injection, are controlled by bubble formation. The understanding of such effects requires therefore the modeling of (1) diffusional reactions of He atoms with other defects; (2) nucleation and growth of He bubbles; (3) transformation of such bubbles into cavities under continuous He generation and irradiation or creep stress. The present paper is focussed on the modeling of the (coupled) high temperature bubble nucleation and growth processes within and on grain boundaries. Two limiting cases are considered: di-atomic nucleation described by the simplest possible sets of rate equations, and multi-atomic nucleation described by classical nucleation theory. Scaling laws are derived which characterize the dependence of the bubble densities upon time (He-dose), He generation rate and temperature. Comparison with experimental data of AISI 316 SS α-implanted at temperatures around 1000 K indicates bubble nucleation of the multi-atomic type. The nucleation and growth models are applied to creep tests performed during α-implantation suggesting that in these cases gas driven bubble growth is the life time controlling mechanism. The narrow (creep stress/He generation rate) range of this mechanism in a mechanism map constructed from these tests indicates that in many reactor situations the time to rupture is probably controlled by stress driven cavity growth rather than by gas driven bubble growth. (orig.)

  8. Transfer of hydrogen and helium through corrugated, flexible tubes

    International Nuclear Information System (INIS)

    Schippl, K.

    2001-01-01

    The transfer of liquid gas or cold gas through corrugated tubes is an alternative to rigid systems for the use in reactor technique. Advantages: flexibility for easy installation; these tubes together with their associated terminations and hardware are assembled, leak-tested and evacuated at the factory. This permits simple and cost saving installation on site. All tubes are helium leak-tested with a sensitivity of 10E -9 mbar 1/sec. Following the leak test, the vacuum space is pumped down to the operation vacuum level and properly sealed. The vacuum integrity is guaranteed as a result of the high degree of cleanliness observed during production and from the use of a specially selected better material inside the vacuum space. Disadvantage: pressure is limited to 20 bar. To fulfil all rules of the reactor safety, different tests have to be done. Because of the longitudinal weld of the corrugated tube, a bursting test of different sizes gives the best information of the liability of this kind of tube. It can be shown that the bursting pressure of such a tube is more than 5 times higher than the max. working pressure

  9. Hydrogen and helium adsorption on potassium

    International Nuclear Information System (INIS)

    Garcia, R.; Mulders, N.; Hess, G.

    1995-01-01

    A previous quartz microbalance study of adsorption of helium on sodium indicates that the inert layer is surprisingly small. Similar experiments with hydrogen on sodium show layer by layer growth below a temperature of 7K. These results motivated the authors to extend the experiments to lower temperatures. A suitable apparatus, capable of reaching 0.45 K, while still enabling them to do in situ alkali evaporation, has been constructed. The authors will report on the results of microbalance adsorption experiments of helium and hydrogen on potassium

  10. Electron-helium scattering in Debye plasmas

    International Nuclear Information System (INIS)

    Zammit, Mark C.; Fursa, Dmitry V.; Bray, Igor; Janev, R. K.

    2011-01-01

    Electron-helium scattering in weakly coupled hot-dense (Debye) plasma has been investigated using the convergent close-coupling method. The Yukawa-type Debye-Hueckel potential has been used to describe plasma Coulomb screening effects. Benchmark results are presented for momentum transfer cross sections, excitation, ionization, and total cross sections for scattering from the ground and metastable states of helium. Calculations cover the entire energy range up to 1000 eV for the no screening case and various Debye lengths (5-100 a 0 ). We find that as the screening interaction increases, the excitation and total cross sections decrease, while the total ionization cross sections increase.

  11. Rotary magnetic refrigerator for superfluid helium production

    International Nuclear Information System (INIS)

    Hakuraku, Y.; Ogata, H.

    1986-01-01

    A new rotary-magnetic refrigerator designed to obtain superfluid helium temperatures by executing a magnetic Carnot cycle is developed. A rotor containing 12 magnetic refrigerants (gadolinium-gallium-garnet) is immersed in liquid helium at 4.2 K and rotated at constant speed in a steady magnetic field distribution. Performance tests demonstrate that the new rotary refrigerator is capable of obtaining a temperature of 1.48 K. The maximum useful cooling power obtained at 1.8 K is 1.81 W which corresponds to a refrigeration efficiency of 34%

  12. Effect of helium irradiation on fracture modes

    International Nuclear Information System (INIS)

    Hanamura, T.; Jesser, W.A.

    1982-01-01

    The objective of this work is to determine the crack opening mode during in-situ HVEM tensile testing and how it is influenced by test temperature and helium irradiation. Most cracks were mixed mode I and II. However, between 250 0 C and room temperature the effect of helium irradiation is to increase the amount of mode I crack propagation. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in irradiated specimens tested between 250 0 C and room temperature, but could be restored by a post irradiation anneal

  13. Calculation of electron-helium scattering

    International Nuclear Information System (INIS)

    Fursa, D.V.; Bray, I.

    1994-11-01

    We present the Convergent Close-Coupling (CCC) theory for the calculation of electron-helium scattering. We demonstrate its applicability at a range of projectile energies of 1.5 to 500 eV to scattering from the ground state to n ≤3 states. Excellent agreement with experiment is obtained with the available differential, integrated, ionization, and total cross sections, as well as with the electron-impact coherence parameters up to and including the 3 3 D state excitation. Comparison with other theories demonstrates that the CCC theory is the only general reliable method for the calculation of electron helium scattering. (authors). 66 refs., 2 tabs., 24 figs

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  15. Options for helium circulation in a hydrogen production plant VHTR-Si: thermal-economic comparative

    International Nuclear Information System (INIS)

    Mendoza A, A.; Francois L, J. L.; Anaya D, A.

    2011-11-01

    The technologies that take advantage of the heat of nuclear reactors of IV generation are of great interest, due to their high energy efficiencies and to their strong economic potential. An example of these technologies is the sulfur-iodine process coupled to a nuclear reactor of high temperature cooled by helium. In this process heat is transferred from the nuclear reactor to the chemical plant by means of two cycles of helium interconnected by an intermediate heat exchanger. The first, denominated primary cycle of cooling, removes the heat of the nuclear reactor, transferring to the secondary cycle to be distributed to equipment s in the chemical plant. The pass of the helium gas through the equipment s that compose each one of the cycles, implies pressure losses that should be compensated necessarily by re-compression to maintain a stable state in the system, causing the energy consumption, usually rejected in the energy analyses. When to this energy is added the energy required in the hydrogen plant: energy required by the pumping systems, will decrease the efficiency of the nucleus-chemical complex, increasing the even cost of the hydrogen. In this work, three options to supply the compression energy and pumping (CEP) to the system are proposed, and these are analyzed thermodynamic and economically. The results indicate that to consider the CEP in the economic analysis increases between 1.5 and 3% the even cost of the hydrogen, and that the option with more energy efficiency is not necessarily the best for the nucleus-chemical complex. (Author)

  16. Dissolved helium and TDS in groundwater from Bhavnagar in Gujarat

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    2003-01-02

    Jan 2, 2003 ... by enhanced pumping of old groundwater with relatively higher concentration of dissolved helium and salt .... solubility changes due to these (Weiss 1971) can- ... aquifers and relatively low helium concentra- .... permeability.

  17. A new helium gas recovery and purification system

    International Nuclear Information System (INIS)

    Yamamotot, T.; Suzuki, H.; Ishii, J.; Hamana, I.; Hayashi, S.; Mizutani, S.; Sanjo, S.

    1974-01-01

    A helium gas recovery and purification system, based on the principle of gas permeation through a membrane, is described. The system can be used for the purification of helium gas containing air as a contaminant. The apparatus, operating at ambient temperature does not need constant attention, the recovery ratio of helium gas is satisfactory and running costs are low. Gases other than helium can be processed with the apparatus. (U.K.)

  18. Helium release from metals with face-centered cubic structure

    International Nuclear Information System (INIS)

    Sciani, V.; Lucki, G.; Jung, P.

    1984-01-01

    The helium release from gold sheets of 5 and 54 μm of thickness and helium concentrations from 10 -9 to 10 -5 ap of He during the isothermal and linear annealing is studied. The helium was put in the sample through the implantation of alpha particles, with variable energy,in the cyclotron. The free diffusion of the atoms of the helium, where the diffusion coefficient follows an Arrhenius law is studied. (E.G.) [pt

  19. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    International Nuclear Information System (INIS)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

  20. Analysis of a 115MW, 3 shaft, helium Brayton cycle

    International Nuclear Information System (INIS)

    Pradeepkumar, K.N.

    2002-01-01

    This research theme is originated from a development project that is going on in South Africa, for the design and construction of a closed cycle gas turbine plant using gas-cooled reactor as the heat source to generate 115 MW of electricity. South African Power utility company, Eskorn, promotes this developmental work through its subsidiary called PBMR (Pebble Bed Modular Reactor). Some of the attractive features of this plant are the inherent and passive safety features, modular geometry, small evacuation area, small infrastructure requirements for the installation and running of the plant, small construction time, quick starting and stopping and also low operational cost. This exercise is looking at the operational aspects of a closed cycle gas turbine, the finding of which will have a direct input towards the successful development and commissioning of the plant. A thorough understanding of the fluid dynamics in this three-shaft system and its transient performance analysis were the two main objectives of this research work. A computer programme called GTSI, developed by a previous Cranfield University research student, has been used in this as a base programme for the performance analysis. Some modifications were done on this programme to improve its control abilities. The areas covered in the performance analysis are Start-up, Shutdown and Load ramping. A detailed literature survey has been conducted to learn from the helium Turbo machinery experiences, though it is very limited. A critical analysis on the design philosophy of the PBMR is also carried out as part of this research work. The performance analysis has shown the advantage, disadvantage and impact of various power modulation methods suggested for the PBMR. It has tracked the effect of the operations of the various valves included in the PBMR design. The start-up using a hot gas injection has been analysed in detail and a successful start region has been mapped. A start-up procedure is also written

  1. Helium-air exchange flows through partitioned opening and two-opening

    International Nuclear Information System (INIS)

    Kang, T. I.

    1997-01-01

    This paper describes experimental investigations of helium-air exchange flows through partitioned opening and two-opening. Such exchange flows may occur following rupture accident of stand pipe in high temperature engineering test reactor. A test vessel with the two types of small opening on top of test cylinder is used for experiments. An estimation method of mass increment is developed to measure the exchange flow rate. Upward flow of the helium and downward flow of the air in partitioned opening system interact out of entrance and exit of the opening. Therefore, an experiment with two-opening system is made to investigate effect of the fluids interaction of partitioned opening system. As a result of comparison of the exchange flow rates between two types of the opening system, it is demonstrated that the exchange flow rate of the two-opening system is larger than that of the partitioned opening system because of absence of the effect of fluids interaction. (author)

  2. Numerical modeling and validation of helium jet impingement cooling of high heat flux divertor components

    International Nuclear Information System (INIS)

    Koncar, Bostjan; Simonovski, Igor; Norajitra, Prachai

    2009-01-01

    Numerical analyses of jet impingement cooling presented in this paper were performed as a part of helium-cooled divertor studies for post-ITER generation of fusion reactors. The cooling ability of divertor cooled by multiple helium jets was analysed. Thermal-hydraulic characteristics and temperature distributions in the solid structures were predicted for the reference geometry of one cooling finger. To assess numerical errors, different meshes (hexagonal, tetra, tetra-prism) and discretisation schemes were used. The temperatures in the solid structures decrease with finer mesh and higher order discretisation and converge towards finite values. Numerical simulations were validated against high heat flux experiments, performed at Efremov Institute, St. Petersburg. The predicted design parameters show reasonable agreement with measured data. The calculated maximum thimble temperature was below the tile-thimble brazing temperature, indicating good heat removal capability of reference divertor design. (author)

  3. High-Temperature Structural Analysis Model of the Process Heat Exchanger for Helium Gas Loop (II)

    International Nuclear Information System (INIS)

    Song, Kee Nam; Lee, Heong Yeon; Kim, Chan Soo; Hong, Seong Duk; Park, Hong Yoon

    2010-01-01

    PHE (Process Heat Exchanger) is a key component required to transfer heat energy of 950 .deg. C generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the helium gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the helium gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype

  4. Calculation of displacement and helium production at the LAMPF irradiation facility

    International Nuclear Information System (INIS)

    Wechsler, M.S.; Davidson, D.R.; Sommer, W.F.; Greenwood, L.R.

    1985-01-01

    Differential and total displacement and helium-production rates are calculated for copper irradiated by spallation neutrons and 760-MeV protons at LAMPF. The calculations are performed using the SPECTOR and VNMTC computer codes, the latter being specially designed for spallation radiation-damage calculations. For comparison, similar SPECTER calculations are also described for irradiation of copper in the experimental breeder reactor (EBR-II) at the Argonne National Laboratory-West in Idaho, and in the rotating target neutron source (RTNS-II) at Lawrence Livermore Laboratory. The neutron energy spectra for LAMPF, EBR-II, and RTNS-II and the displacement and helium-production cross sections are shown

  5. 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)

  6. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    1988-01-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  7. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

  8. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  9. Construction and performance tests of Helium Engineering Demonstration Loop (HENDEL) for VHTR

    International Nuclear Information System (INIS)

    Hishida, M.; Tanaka, T.; Shimomura, H.; Sanokawa, K.

    1984-01-01

    A helium engineering demonstration loop (HENDEL) was constructed and operated in JAERI in order to develop the high-temperature key components of an experimental very high temperature gas cooled reactor, like fuel stack, in-core reactor structure, hot gas duct, intermediate heat exchanger. Performance tests as well as demonstration of integrity are carried out with large-size or actual-size models of key components. The key components to be tested in HENDEL are: fuel stack and control rod; core supporting structure, or bottom structure of rector core exposed to direct impingement of high temperature core outlet flow; reactor internal components and structure; high temperature components in heat removal system (primary and secondary cooling systems)

  10. Nuclear fuel rod helium leak inspection apparatus and method

    International Nuclear Information System (INIS)

    Ahmed, H.J.

    1991-01-01

    This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod

  11. Observation of visible emission from the molecular helium ion in the afterglow of a dense helium Z-pinch plasma

    International Nuclear Information System (INIS)

    Tucker, J.E.; Brake, M.L.; Gilgenbach, R.M.

    1986-01-01

    The authors present the results of axial and radial time resolved visible emission spectroscopy from the afterglow of a dense helium Z-pinch. These results show that the visible emissions in the pinch afterglow are dominated by line emissions from molecular helium and He II. Axial spectroscopy measurements show the occurrence of several absorption bands which cannot be identified as molecular or atomic helium nor impurities from the discharge chamber materials. The authors believe that these absorption bands are attributable to the molecular helium ion which is present in the discharge. The molecular ion has been observed by others in low pressure and temperature helium discharges directly by means of mass spectrometry and indirectly by the presence of helium atoms in the 2/sup 3/S state, (the He 2/sup 3/S state is believed to result from molecular helium ion recombination). However, the molecular helium ion has not previously been observed spectroscopically

  12. Cooling by mixing of helium isotopes

    International Nuclear Information System (INIS)

    Hansen, O.P.; Olsen, M.; Rasmussen, F.B.

    1975-01-01

    The principles of the helium dilution refrigerator are outlined. The lowest temperature attained with a continuously operated dilution refrigerator was about 10 mK, and 5 mK for a limited period when the supply of concentrated 3 He to the mixing chamber was interrupted. (R.S.)

  13. Diffusion of helium in the Sun

    Energy Technology Data Exchange (ETDEWEB)

    Noerdlinger, P D [Michigan State Univ., East Lansing (USA). Dept. of Astronomy and Astrophysics; Amsterdam Univ. (Netherlands). Sterrenkundig Instituut)

    1977-05-01

    I have reduced the set of diffusion and flow equations developed by Burgers for a multi-component gas to a workable scheme for the actual evaluation of the relative diffusion of hydrogen and helium in stars. Previous analyses have used the Aller and Chapman equations, which apply only to trace constitutents and whose coefficients are not believed to be as accurate as Burgers'. Furthermore, the resulting equations have been combined consistently with Paczynski's stellar evolution code to demonstrate small but significant effects in the Sun, from the thermal and gravitational settling of Helium. The core helium content of a 1 M star goes up about 0.04 and the surface helium content down by about -0.03 in 4.5 10/sup 9/ years. The results are still somewhat uncertain because of uncertainties in the underlying plasma physics, and further research is suggested. In any case, the diffusion process speeds up with time, due to increased temperature gradient, and it will be of interest to follow the process in older stars and in later stellar evolution.

  14. Sounds in one-dimensional superfluid helium

    International Nuclear Information System (INIS)

    Um, C.I.; Kahng, W.H.; Whang, E.H.; Hong, S.K.; Oh, H.G.; George, T.F.

    1989-01-01

    The temperature variations of first-, second-, and third-sound velocity and attenuation coefficients in one-dimensional superfluid helium are evaluated explicitly for very low temperatures and frequencies (ω/sub s/tau 2 , and the ratio of second sound to first sound becomes unity as the temperature decreases to absolute zero

  15. Tritium and helium-3 in metals

    International Nuclear Information System (INIS)

    Lasser, R.

    1989-01-01

    The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

  16. Atmospheric helium and geomagnetic field reversals.

    Science.gov (United States)

    Sheldon, W. R.; Kern, J. W.

    1972-01-01

    The problem of the earth's helium budget is examined in the light of recent work on the interaction of the solar wind with nonmagnetic planets. It is proposed that the dominant mode of helium (He4) loss is ion pumping by the solar wind during geomagnetic field reversals, when the earth's magnetic field is very small. The interaction of the solar wind with the earth's upper atmosphere during such a period is found to involve the formation of a bow shock. The penetration altitude of the shock-heated solar plasma is calculated to be about 700 km, and ionization rates above this level are estimated for a cascade ionization (electron avalanche) process to average 10 to the 9th power ions/sq cm/sec. The calculated ionization rates and the capacity of the solar wind to remove ionized helium (He4) from the upper atmosphere during geomagnetic dipole reversals are sufficient to yield a secular equilibrium over geologic time scales. The upward transport of helium from the lower atmosphere under these conditions is found to be adequate to sustain the proposed loss rate.

  17. NUCLEAR CONDENSATE AND HELIUM WHITE DWARFS

    Energy Technology Data Exchange (ETDEWEB)

    Bedaque, Paulo F.; Berkowitz, Evan [Maryland Center for Fundamental Physics, Department of Physics, University of Maryland, College Park, MD (United States); Cherman, Aleksey, E-mail: bedaque@umd.edu, E-mail: evanb@umd.edu, E-mail: a.cherman@damtp.cam.ac.uk [Department of Applied Mathematics and Theoretical Physics, University of Cambridge, Cambridge CB3 0WA (United Kingdom)

    2012-04-10

    We consider a high-density region of the helium phase diagram, where the nuclei form a Bose-Einstein condensate rather than a classical plasma or a crystal. Helium in this phase may be present in helium-core white dwarfs. We show that in this regime there is a new gapless quasiparticle not previously noticed, arising when the constraints imposed by gauge symmetry are taken into account. The contribution of this quasiparticle to the specific heat of a white dwarf core turns out to be comparable in a range of temperatures to the contribution from the particle-hole excitations of the degenerate electrons. The specific heat in the condensed phase is two orders of magnitude smaller than in the uncondensed plasma phase, which is the ground state at higher temperatures, and four orders of magnitude smaller than the specific heat that an ion lattice would provide, if formed. Since the specific heat of the core is an important input for setting the rate of cooling of a white dwarf star, it may turn out that such a change in the thermal properties of the cores of helium white dwarfs has observable implications.

  18. Helium ion lithography principles and performance

    NARCIS (Netherlands)

    Drift, E. van der; Maas, D.J.

    2012-01-01

    Recent developments show that Scanning Helium Ion Beam Lithography (SHIBL) with a sub-nanometer beam diameter is a promising alternative fabrication technique for high-resolution nanostructures at high pattern densities. Key principles and critical conditions of the technique are explained. From

  19. Resource letter SH-1: superfluid helium

    International Nuclear Information System (INIS)

    Hallock, R.B.

    1982-01-01

    The resource letter covers the general subject of superfluid helium and treats 3 He and 3 He-- 4 He mixtures as well as 4 He. No effort has been made to include the fascinating experiments on either solid helium or the equally fascinating work on adsorbed helium where the helium coverage is below that necessary for superfluidity. An earlier resource letter by C. T. Lane [Am. J. Phys. 35, 367 (1967)] may be consulted for additional comments on some of the cited earlier manuscripts, but the present work is self-contained and may be used independently. Many high-quality research reports have not been cited here. Rather, the author has tried in most cases to include works particularly readable or relevant. There is a relatively heavy emphasis on experimental references. The primary reason is that these works tend to be more generally readable. No doubt some works that might have been included, have not, and for this the author takes responsibility with apology. Articles selected for incorporation in a reprint volume (to be published separately by the American Association of Physics Teachers) are marked with an asterisk(*). Following each referenced work the general level of difficulty is indicated by E, I, or A for elementary, intermediate, or advanced

  20. Photoassociation of cold metastable helium atoms

    NARCIS (Netherlands)

    Woestenenk, G.R.

    2001-01-01

    During the last decades the study of cold atoms has grown in a great measure. Research in this field has been made possible due to the development of laser cooling and trapping techniques. We use laser cooling to cool helium atoms down to a temperature of 1 mK and we are able to