WorldWideScience

Sample records for metal moderated reactors

  1. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Energy Technology Data Exchange (ETDEWEB)

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  2. Reactor-moderated intermediate-energy neutron beams for neutron-capture therapy

    International Nuclear Information System (INIS)

    Less, T.J.

    1987-01-01

    One approach to producing an intermediate energy beam is moderating fission neutrons escaping from a reactor core. The objective of this research is to evaluate materials that might produce an intermediate beam for NCT via moderation of fission neutrons. A second objective is to use the more promising moderator material in a preliminary design of an NCT facility at a research reactor. The evaluations showed that several materials or combinations of materials could produce a moderator source for an intermediate beam for NCT. The best neutron spectrum for use in NCT is produced by Al 2 O 3 , but mixtures of Al metal and D 2 O are also attractive. Using the best moderator materials, results were applied to the design of an NCT moderator at the Georgia Institute of Technology Research Reactor's bio-medical facility. The amount of photon shielding and thermal neutron absorber were optimized with respect to the desired photon dose rate and intermediate neutron flux at the patient position

  3. Quality assurance in the manufacture of metallic uranium fuel for research reactors

    International Nuclear Information System (INIS)

    Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

    2009-01-01

    Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

  4. Moderator for nuclear reactor

    International Nuclear Information System (INIS)

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  5. Power spectral density measurements with 252Cf for a light water moderated research reactor

    International Nuclear Information System (INIS)

    King, W.T.; Mihalczo, J.T.

    1979-01-01

    A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

  6. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  7. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  8. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  9. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  10. 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.)

  11. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

    In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

  12. Criticality evaluations with moderators other than water for uranium metal fuels

    International Nuclear Information System (INIS)

    Toffer, H.; Tollefson, D.A.; Finfrock, S.H.

    1986-01-01

    Occasionally, nuclear criticality safety analyses of fissile material handling operations or transport situations require consideration of moderation other than water. Such moderators could be oils, plastics, wood, concrete, carbon, or even wet sand. All of these materials contain either hydrogen, carbon, or mixtures of the two elements as the principal moderators. Other elements as part of the compounds or mixtures contribute less to the neutron slowing down process and can possibly be significant parasitic neutron absorbers. Results of a series of calculations are presented illustrating the impact of various moderators on critical masses or critical parameters as a function of lattice pitch for different uranium metal fuel elements at low 235 U enrichments. Several nuclear criticality safety analyses performed at the Hanford N Reactor, operated by UNC Nuclear Industries for the US Department of Energy, have considered alternative moderators to assure that water moderation represented the most limiting case

  13. 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.)

  14. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

    Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

  15. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  16. Study on the effect of moderator density reactivity for Kartini reactor

    International Nuclear Information System (INIS)

    Budi Rohman; Widarto

    2009-01-01

    One of important characteristics of water-cooled reactors is the change of reactivity due to change in the density of coolant or moderator. This parameter generally has negative value and it has significant role in preventing the excursion of power during operation. Many thermal-hydraulic codes for nuclear reactors require this parameter as the input to account for reactivity feedback due to increase in moderator voids and the subsequent decrease in moderator density during operation. Kartini reactor is cooled and moderated by water, therefore, it is essential to study the effect of the change in moderator density as well as to determine the value of void or moderator density reactivity coefficient in order to characterize its behavior resulting from the presence of vapor or change of moderator density during operation. Analysis by MCNP code shows that the reactivity of core is decreasing with the decrease in moderator density. The analysis estimates the void or moderator density reactivity coefficient for Kartini Reactor to be -2.17×10-4 Δρ/ % void . (author)

  17. Measuring device for the temperature coefficient of reactor moderators

    International Nuclear Information System (INIS)

    Nakano, Yuzo.

    1987-01-01

    Purpose: To rapidly determine by automatic calculation the temperature coefficient for moderators which has been determined so far by a log of manual processings. Constitution: Each of signals from a control rod position indicator, a reactor reactivity, instrument and moderator temperature meter are inputted, and each of the signals and designed valued for the doppler temperature coefficients are stored. Recurling calculation is conducted based on the reactivity and the moderator temperature at an interval where the temperature changes of the moderators are equalized at an identical control rod position, to determine isothermic coefficient. Then, the temperature coefficient for moderator are calculated from the isothermic coefficient and the doppler temperature coefficient. The relationship between the reactivity and the moderator temperature is plotted on a X-Y recorder. The stored signals and the calculated temperature coefficient for moderators are sequentially displayed and the results are printed out when the measurement is completed. According to the present device, since the real time processing is conducted, the processing time can be shortened remarkably. Accordingly, it is possible to save the man power for the test of the nuclear reactor and improve the reactor operation performance. (Kamimura, M.)

  18. The high moderating ratio reactor using 100% MOX reloads

    International Nuclear Information System (INIS)

    Barbrault, P.

    1994-06-01

    This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

  19. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  20. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

    Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

    1997-12-31

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

  1. Fluid moderator control system reactor internals distribution system

    International Nuclear Information System (INIS)

    Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

    1987-01-01

    This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

  2. Evolution of the liquid metal reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    This paper reports on the integral fat reactor (IFR) concept. A key feature of the IFR concept is the metallic fuel, the original choice in liquid metal reactor development. An IFR development program is detailed by the authors

  3. Status and perspective of development of cold moderators at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  4. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  5. Summary of the 4th workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  6. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  7. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  8. Summary of the 4th workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  9. The US Liquid Metal Reactor Development Program

    International Nuclear Information System (INIS)

    Till, C.E.; Arnold, W.H.; Griffith, J.D.

    1988-01-01

    The US Liquid Metal Reactor Development Program has been restructured to take advantage of the opportunity today to carry out R and D on truly advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the Integral Fast Reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable US program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and, with international cooperation, aqueous reprocessing. Design studies are carried out in conjunction with the IFR technology development program. In summary, the US maintains an active development program in Liquid Metal Reactor technology, and new directions in reactor safety are central to the program

  10. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

  11. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1984-01-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  12. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  13. Solid methane cold moderator for the IBR-2 reactor

    International Nuclear Information System (INIS)

    Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

    1997-09-01

    The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

  14. FLUID MODERATED REACTOR

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  15. Evaluation of Metal-Fueled Surface Reactor Concepts

    International Nuclear Information System (INIS)

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-01

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

  16. Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

    2017-01-01

    Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

  17. Summary of the 3rd workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  18. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  19. Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1990-01-01

    The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

  20. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

  1. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  2. Physical particularities of nuclear reactors using heavy moderators of neutrons

    International Nuclear Information System (INIS)

    Kulikov, G. G.; Shmelev, A. N.

    2016-01-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  3. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  4. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  5. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  6. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  7. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon

    2015-01-01

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW th and an electricity generation mode of 100 kW th , equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics

  8. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Directory of Open Access Journals (Sweden)

    Seung Hyun Nam

    2015-10-01

    Full Text Available Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER, for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MWth and an electricity generation mode of 100 kWth, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  9. Conceptual design of a pressure tube light water reactor with variable moderator control

    International Nuclear Information System (INIS)

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-01-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  10. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  11. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  12. Development of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Park, C.K.

    1998-01-01

    Future nuclear power plants should not only have the features of improved safety and economic competitiveness but also provide a means to resolve spent fuel storage problems by minimizing volume of high level wastes. It is widely believed that liquid metal reactors (LMRs) have the highest potential of meeting these requirements. In this context, the LMR development program was launched as a national long-term R and D program in 1992, with a target to introduce a commercial LMR around 2030. Korea Advanced Liquid Metal Reactor (KALIMER), a 150 MWe pool-type sodium cooled prototype reactor, is currently under the conceptual design study with the target schedule to complete its construction by the mid-2010s. This paper summarizes the KALIMER development program and major technical features of the reactor system. (author)

  13. Power control device for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

    Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

  14. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  15. Reactivity margins in heavy water moderated production reactors

    International Nuclear Information System (INIS)

    Benton, F.D.

    1981-11-01

    The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

  16. Copper extraction from coarsely ground printed circuit boards using moderate thermophilic bacteria in a rotating-drum reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Michael L.M., E-mail: mitchel.marques@yahoo.com.br [Bio& Hydrometallurgy Laboratory, Department of Metallurgical and Materials Engineering, Universidade Federal de Ouro Preto, Campus Morro do Cruzeiro, Ouro Preto, MG 35400-000 (Brazil); Leão, Versiane A., E-mail: versiane@demet.em.ufop.br [Bio& Hydrometallurgy Laboratory, Department of Metallurgical and Materials Engineering, Universidade Federal de Ouro Preto, Campus Morro do Cruzeiro, Ouro Preto, MG 35400-000 (Brazil); Gomes, Otavio [Centre for Mineral Technology – CETEM, Av Pedro Calmon, 900, 21941-908 Rio de Janeiro (Brazil); Lambert, Fanny; Bastin, David; Gaydardzhiev, Stoyan [Mineral Processing and Recycling, University of Liege, SartTilman, 4000 Liege (Belgium)

    2015-07-15

    Highlights: • Copper bioleaching from PCB (20 mm) by moderate thermophiles was demonstrated. • Larger PCB sheets enable a cost reduction due to the elimination of fine grinding. • Crushing generated cracks in PCB increasing the copper extraction. • A pre-treatment step was necessary to remove the lacquer coating. • High copper extractions (85%) were possible with pulp density of up to 25.0 g/L. - Abstract: The current work reports on a new approach for copper bioleaching from Printed Circuit Board (PCB) by moderate thermophiles in a rotating-drum reactor. Initially leaching of PCB was carried out in shake flasks to assess the effects of particle size (−208 μm + 147 μm), ferrous iron concentration (1.25–10.0 g/L) and pH (1.5–2.5) on copper leaching using mesophile and moderate thermophile microorganisms. Only at a relatively low solid content (10.0 g/L) complete copper extraction was achieved from the particle size investigated. Conversely, high copper extractions were possible from coarse-ground PCB (20 mm-long) working with increased solids concentration (up to 25.0 g/L). Because there was as the faster leaching kinetics at 50 °C Sulfobacillus thermosulfidooxidans was selected for experiments in a rotating-drum reactor with the coarser-sized PCB sheets. Under optimal conditions, copper extraction reached 85%, in 8 days and microscopic observations by SEM–EDS of the on non-leached and leached material suggested that metal dissolution from the internal layers was restricted by the fact that metal surface was not entirely available and accessible for the solution in the case of the 20 mm-size sheets.

  17. Copper extraction from coarsely ground printed circuit boards using moderate thermophilic bacteria in a rotating-drum reactor

    International Nuclear Information System (INIS)

    Rodrigues, Michael L.M.; Leão, Versiane A.; Gomes, Otavio; Lambert, Fanny; Bastin, David; Gaydardzhiev, Stoyan

    2015-01-01

    Highlights: • Copper bioleaching from PCB (20 mm) by moderate thermophiles was demonstrated. • Larger PCB sheets enable a cost reduction due to the elimination of fine grinding. • Crushing generated cracks in PCB increasing the copper extraction. • A pre-treatment step was necessary to remove the lacquer coating. • High copper extractions (85%) were possible with pulp density of up to 25.0 g/L. - Abstract: The current work reports on a new approach for copper bioleaching from Printed Circuit Board (PCB) by moderate thermophiles in a rotating-drum reactor. Initially leaching of PCB was carried out in shake flasks to assess the effects of particle size (−208 μm + 147 μm), ferrous iron concentration (1.25–10.0 g/L) and pH (1.5–2.5) on copper leaching using mesophile and moderate thermophile microorganisms. Only at a relatively low solid content (10.0 g/L) complete copper extraction was achieved from the particle size investigated. Conversely, high copper extractions were possible from coarse-ground PCB (20 mm-long) working with increased solids concentration (up to 25.0 g/L). Because there was as the faster leaching kinetics at 50 °C Sulfobacillus thermosulfidooxidans was selected for experiments in a rotating-drum reactor with the coarser-sized PCB sheets. Under optimal conditions, copper extraction reached 85%, in 8 days and microscopic observations by SEM–EDS of the on non-leached and leached material suggested that metal dissolution from the internal layers was restricted by the fact that metal surface was not entirely available and accessible for the solution in the case of the 20 mm-size sheets

  18. Neutron absorption profile in a reactor moderated by different mixtures of light and heavy waters

    International Nuclear Information System (INIS)

    Nagy, Mohamed E.; Aly, Mohamed N.; Gaber, Fatma A.; Dorrah, Mahmoud E.

    2014-01-01

    Highlights: • We studied neutron absorption spectra in a mixed water moderated reactor. • Changing D 2 O% in moderator induced neutron energy spectral shift. • Most of the neutrons absorbed in control rods were epithermal. • Control rods worth changes were not proportional to changes of D 2 O% in moderator. • Control rod arrangement influenced the neutronic behavior of the reactor. - Abstract: A Monte-Carlo parametric study was carried out to investigate the neutron absorption profile in a model of LR-0 reactor when it is moderated by different mixtures of heavy/light waters at molecular ratios ranging from 0% up to 100% D 2 O at increments of 10% in D 2 O. The tallies included; neutron absorption profiles in control rods and moderator, and neutron capture profile in 238 U. The work focused on neutron absorption in control rods entailing; total mass of control rods needed to attain criticality, neutron absorption density and total neutron absorption in control rods at each of the studied mixed water moderators. The aim was to explore whether thermal neutron poisons are the most suitable poisons to be used in control rods of nuclear reactors moderated by mixed heavy/light water moderators

  19. Small liquid metal reactor for an initial phase of fast breeder reactor introduction

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1985-01-01

    Safety and burnup characteristics of a 1000 MWth liquid metal reactor have been examined for various fuel types. With metallic Pu/Th fuel containing a small amount of zirconium hydride, low sodium-void reactivity, a high Doppler coefficient, and small burnup reactivity swings can be achieved. A conservative design is considered for an initial phase of fast breeder reactor development and possible modifications are discussed. (Author) [pt

  20. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  1. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    Pinedo V, J.L.

    1979-01-01

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  2. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2000-01-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  3. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  4. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  5. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  6. The US Liquid-Metal Reactor Program - overview and status

    International Nuclear Information System (INIS)

    Quinn, J.E.; Gyorey, G.L.; Salerno, L.N.

    1992-01-01

    The US Advanced Liquid-Metal Reactor (ALMR) Program has three major elements being developed in an integrated fashion to produce a system meeting the US long-term nuclear energy needs. Reactor design, one of those elements, is the focus of this paper. The other two elements, the integral fast reactor metal-fuel cycle and the light water reactor (LWR) spent-fuel actinide recycle, will be addressed in companion papers. The ALMR is adaptable to multiple missions with few modifications such as the core arrangements. The missions identified to date are (a) the extension of the existing uranium resources through breeding and highly efficient uranium utilization, (b) the recycle and utilization of the long-life actinides in LWR spent fuel as fissile material for the ALMR, and (c) the conversion of excess weapons fissil material into electricity. In addition to these missions, the reactor design is adaptable to either the metal-fuel cycle or the oxide fuel cycle

  7. Device for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Yutaro; Kawarazaki, Yuki; Sugiyama, Yu.

    1996-01-01

    A member comprising hydrogen occluding materials is introduced to a reactor incorporated with U-235 as fuels in order to moderate and breed fast neutrons and to control the reactor. Since the amount of light hydrogen or heavy hydrogen is substantially the same as that of metal, etc. of hydrogen occluding material, a moderating efficiency substantially equal with that of a moderator comprising H 2 O can be obtained. In addition, since the member acting as a moderator has an effect of multiplying neutrons, use of only natural uranium 0.72% as nuclear fuels causes chain reaction to provide a function as a nuclear reactor. Further, the hydrogen occluding material can be used also as a control rod for controlling the reactor. The hydrogen occluding material may be Ti, Zr, Pd, proton conductor, Ag, Pt, Rh or oxides thereof or alloys thereof. The member comprising hydrogen occluding materials is preferably coated with a material not permeating hydrogen. (N.H.)

  8. Selective removal of heavy metals from metal-bearing wastewater in a cascade line reactor.

    Science.gov (United States)

    Pavlović, Jelena; Stopić, Srećko; Friedrich, Bernd; Kamberović, Zeljko

    2007-11-01

    This paper is a part of the research work on 'Integrated treatment of industrial wastes towards prevention of regional water resources contamination - INTREAT' the project. It addresses the environmental pollution problems associated with solid and liquid waste/effluents produced by sulfide ore mining and metallurgical activities in the Copper Mining and Smelting Complex Bor (RTB-BOR), Serbia. However, since the minimum solubility for the different metals usually found in the polluted water occurs at different pH values and the hydroxide precipitates are amphoteric in nature, selective removal of mixed metals could be achieved as the multiple stage precipitation. For this reason, acid mine water had to be treated in multiple stages in a continuous precipitation system-cascade line reactor. All experiments were performed using synthetic metal-bearing effluent with chemical a composition similar to the effluent from open pit, Copper Mining and Smelting Complex Bor (RTB-BOR). That effluent is characterized by low pH (1.78) due to the content of sulfuric acid and heavy metals, such as Cu, Fe, Ni, Mn, Zn with concentrations of 76.680, 26.130, 0.113, 11.490, 1.020 mg/dm3, respectively. The cascade line reactor is equipped with the following components: for feeding of effluents, for injection of the precipitation agent, for pH measurements and control, and for removal of the process gases. The precipitation agent was 1M NaOH. In each of the three reactors, a changing of pH and temperature was observed. In order to verify. efficiency of heavy metals removal, chemical analyses of samples taken at different pH was done using AES-ICP. Consumption of NaOH in reactors was 370 cm3, 40 cm3 and 80 cm3, respectively. Total time of the experiment was 4 h including feeding of the first reactor. The time necessary to achieve the defined pH value was 25 min for the first reactor and 13 min for both second and third reactors. Taking into account the complete process in the cascade line

  9. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  10. A large economic liquid metal reactor for United States utilities

    International Nuclear Information System (INIS)

    Rodwell, E.

    1985-01-01

    The United States has demonstrated its ability to build and operate small and medium sized liquid metal reactors and continues to operate the Experimental Breeder Reactor II and the Fast Flux Test Facility to demonstrate long life fuel designs. Similar-sized liquid metal reactors in Europe have been followed by a step-up to the 1200 MWe capacity of the Superphenix plant. To permit the United States to make a similar step-up in capacity, a 1320 MWe liquid metal reactor plant has been designed with the main emphasis on minimizing the specific capital cost in order to be competitive with light water reactor plant and fossil plant alternatives. The design is based on a four parallel heat transport loops arrangement and complies with current regulatory requirements. The primary heat transport loops are now being integrated into the reactor vessel to achieve further reduction in the capital cost

  11. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F.

    2015-01-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW e and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k eff = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  12. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  13. Studies on gadolinium precipitation in moderator system of nuclear reactor

    International Nuclear Information System (INIS)

    Joshi, Akhilesh C.; Rajesh, Puspalata; Rufus, A.L.; Velmurugan, S.

    2015-01-01

    Gadolinium is used in the moderator system of many Pressurised Heavy Water Reactors (PHWRs) for start-up, shut-down and reactivity control during operation. It is very much essential to maintain gadolinium concentration in the system as desired. It has been reported that gadolinium gets precipitated in as oxalate in carbonated water under the influence of γ-radiation. Hence, studies were carried out to investigate the effect of dose, presence of other metal ions and metal surfaces on the precipitation of gadolinium. The results showed that the amount of carboxylic acids viz., formic acid and oxalic acid, formed due to radiolysis is dependent on the dose and that the curve passes though a maxima. Gadolinium is added in higher concentration in Advanced Heavy Water Reactor. So, experiments with high concentration of gadolinium were also carried out. Ultra pure water saturated with high purity CO 2 containing gadolinium and desired ion/surface was irradiated with γ-radiation from 60 Co source at 25°C to doses ranging from 2.5-16.6 Mrad. At lower doses, formation of carboxylic acids takes place but as the dose increases, decomposition of these acids starts and hence the concentration Vs dose passes through a maximum. It was found that precipitation of gadolinium as oxalate occurred at lower doses. At higher doses, it was seen that pH of the solution decreases and hence solubility of gadolinium oxalate increases. It was also observed that the amount of gadolinium precipitated varied linearly with the initial concentration of gadolinium varying from 2 ppm to 20 ppm. While for gadolinium concentration from 20 ppm to 400 ppm, gadolinium in particulate form was observed. The amount of carboxylic acids formed depends on the nature of cations present in solution. It was found that the amount of oxalic acid formed in the case of gadolinium was more than that formed in the case of sodium. Presence of metal oxides such as ZrO 2 formed over zircoloy surfaces was found to

  14. Parametric study of moderator heat exchanger for Candu 6 advanced reactor

    International Nuclear Information System (INIS)

    Umar, Efrizon; Vecchiarelli, Jack

    2000-01-01

    The passive moderator system for Candu 6 advanced reactor require moderator heat exchanger with the small size and the low resistance coefficient of the shell-side. The study is to determine the required size of moderator heat exchanger, and to calculate the shell side of resistance coefficient have been done. Using computer code CATHENA, it is concluded that the moderator heat exchanger can be used at full power-normal operation condition, especially for the cases with 3600 to 8100 number of tube and 15.90 mm tube diameter. This study show that the proposed moderator heat exchanger have given satisfactory results

  15. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  16. Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

    International Nuclear Information System (INIS)

    Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

    1995-01-01

    This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

  17. Developing remote techniques for liquid metal reactors

    International Nuclear Information System (INIS)

    Fenemore, Peter

    1987-01-01

    Three devices have been designed in Britain to meet the need for special remote equipment and techniques required to inspect the reactor vessel and internals of liquid metal reactors. The ''Links Manipulator Under-Sodium Viewing System'' - a device to be used for the surveillance of reactor internals, which are submerged in sodium. An ''Automatic Guided Vehicle'' - a free roving vehicle to be used to survey the externals of the reactor vessel. The ''Snake Manipulator'' - an articulated arm used to gain access to restricted areas. (author)

  18. Report on generation IV technical working group 3 : liquid metal reactors

    International Nuclear Information System (INIS)

    Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

    2002-01-01

    This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

  19. UK methods for studying fuel management in water moderated reactors

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1970-10-01

    Current UK methods for studying fuel management and for predicting the reactor physics performance for both light and heavy water moderated power reactors are reviewed. Brief descriptions are given of the less costly computer codes used for initial assessment studies, and also the more elaborate programs associated with detailed evaluation are discussed. Some of the considerations influencing the accuracy of predictions are included with examples of various types of experimental confirmation. (author)

  20. Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

    International Nuclear Information System (INIS)

    Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

    2014-01-01

    In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

  1. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C.

    2017-01-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO_2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D_2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α"M_T(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  2. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

  3. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  4. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  5. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  6. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  7. Nuclear calculation methods for light water moderated reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1961-02-01

    This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

  8. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

    1991-08-01

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  9. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  10. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  11. Metal burning in graphite-moderated reactors

    International Nuclear Information System (INIS)

    Wichner, R.P.; Ball, S.J.; Daw, C.S.; Thomas, J.F.

    1997-01-01

    Pinto beans, sweet corn, and zucchini squash (Cucurbita pepo var. black beauty) were grown in a randomized complete-block field/pot experiment at a site that contained the highest observed levels of surface gross gamma radioactivity within Los Alamos Canyon (LAC) at Los Alamos National Laboratory. Soils as well as washed edible and nonedible crop tissues were analyzed for various radionuclides and heavy metals. Most radionuclides, with the exception of 3 H and tot U, in soil from LAC were detected in significantly higher concentrations (p -1 . This dose was below the International Commission on Radiological Protection permissible dose limit (PDL) of 100 mrem y -1 from all pathways; however, the addition of other internal and external exposure route factors may increase the overall dose over the PDL. Also, the risk of an excess cancer fatality, based on 74 mrem y -1 , was 3.7 x 10 -5 (37 in a million), which is above the Environmental Protection Agency's (acceptable) guideline of one in a million. 25 refs

  12. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

    2013-01-01

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

  13. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

  14. US/DOE Man-Machine Integration program for liquid metal reactors

    International Nuclear Information System (INIS)

    D'Zmura, A.P.; Seeman, S.E.

    1985-03-01

    The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

  15. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  16. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  17. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    International Nuclear Information System (INIS)

    Monsler, M.J.; Meier, W.R.

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate

  18. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Monsler, M.J.; Meier, W.R.

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

  19. Liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Filonov, V.S.; Zaitsev, B.I.; Artemiev, L.N.; Rakhimov, V.V.

    1976-01-01

    A liquid-metal-cooled reactor is described comprising two rotatable plugs, one of them, having at least one hole, being arranged internally of the other, a recharging mechanism with a guide tube adapted to be moved through the hole of the first plug by means of a drive, and a device for detecting stacks with leaky fuel elements, the recharging mechanism tube serving as a sampler

  20. Fissile fuel assembly for a sub-moderated nuclear reactor

    International Nuclear Information System (INIS)

    Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

    1983-01-01

    Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

  1. Structure optimization of CFB reactor for moderate temperature FGD

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

    2013-07-01

    The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

  2. Nitrogen Removal by Anammox Biofilm Column Reactor at Moderately Low Temperature

    Directory of Open Access Journals (Sweden)

    Tuty Emilia Agustina

    2017-10-01

    Full Text Available The anaerobic ammonium oxidation (anammox as a new biological approach for nitrogen removal has been considered to be more cost-effective compared with the combination of nitrification and denitrification process. However, the anammox bioreactors are mostly explored at high temperature (>300C in which temperature controlling system is fully required. This research was intended to develop and to apply anammox process for high nitrogen concentration removal at ambient temperature used for treating wastewater in tropical countries. An up-flow biofilm column reactor, which the upper part constructed with a porous polyester non-woven fabric material as a carrier to attach the anammox bacteria was operated without heating system. A maximum nitrogen removal rate (NRR of 1.05 kg-N m3 d-1 was reached in the operation days of 178 with a Total Nitrogen (TN removal efficiency of 74%. This showed the biofilm column anammox reactor was successfully applied to moderate high nitrogen removal from synthetic wastewater at moderately low temperature. Keywords: Anammox, biofilm column reactor, ambient temperature, nitrogen removal

  3. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-01-01

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  4. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  5. Core design study on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

    2002-01-01

    The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

  6. Aging of metal components in US nuclear reactors

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Strosnider, J.R.

    1998-01-01

    This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

  7. An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Watanabe, N.

    1976-01-01

    In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

  8. A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

    The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

  9. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

    2012-01-01

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  10. Safety aspects of LMR [liquid metal-cooled reactor] core design

    International Nuclear Information System (INIS)

    Cahalan, J.E.

    1986-01-01

    Features contributing to increased safety margins in liquid metal-cooled reactor (LMR) design are identified. The technical basis is presented for the performance of a pool-type reactor system with an advanced metallic alloy fuel in unprotected accidents. Results are presented from analyses of anticipated transients without scram, including loss-of-flow (LOF), transient overpower (TOP), and loss-of-heat-sink (LOHS) accidents

  11. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  12. Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

    1998-01-01

    The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

  13. Compact power reactor

    International Nuclear Information System (INIS)

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  14. Application of probabilistic risk assessment to advanced liquid metal reactor designs

    International Nuclear Information System (INIS)

    Carroll, W.P.; Temme, M.I.

    1987-01-01

    The United States Department of Energy (US DOE) has been active in the development and application of probabilistic risk assessment methods within its liquid metal breeder reactor development program for the past eleven years. These methods have been applied to comparative risk evaluations, the selection of design features for reactor concepts, the selection and emphasis of research and development programs, and regulatory discussions. The application of probabilistic methods to reactors which are in the conceptual design stage presents unique data base, modeling, and timing challenges, and excellent opportunities to improve the final design. We provide here the background and insights on the experience which the US DOE liquid metal breeder reactor program has had in its application of probabilistic methods to the Clinch River Breeder Reactor Plant project, the Conceptual Design State of the Large Development Plant, and updates on this design. Plans for future applications of probabilistic risk assessment methods are also discussed. The US DOE is embarking on an innovative design program for liquid metal reactors. (author)

  15. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  16. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  17. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  18. A low cost liquid metal reactor design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Anderson, C.A.; Mangus, J.D.

    1984-01-01

    A new, compact Liquid Metal Reactor (LMR) plant arrangement designed by Westinghouse, featuring factory-fabricated modules and an integrated fuel cycle facility, has made it possible to project a commercially competitive LMR plant for the near future. This innovative liquid metal-cooled plant design will allow a combination of capital, fuel, operation and maintenance costs that could be lower than today's fossil-fueled or light water reactor plant costs, and incorporate features which enhance public safety even beyond current high standards. Following early core loadings, the plant feeds only on depleted uranium. No shipment of fuel is required. And the plant can be tailored to produce enough plutonium to meet its need or to provide fuel for other nuclear plants

  19. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  20. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  1. Status of research and development on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Iwamura, Takamichi

    2002-01-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  2. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  3. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  4. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  5. Design and development of rolled joint for moderator sparger channel of an Indian Pressurised Heavy Water Reactor

    International Nuclear Information System (INIS)

    Joemon, V.; Sinha, R.K.

    1993-01-01

    Indian Pressurised Heavy Water Reactors are natural uranium fuelled heavy water moderated and cooled reactors. As per the conventional scheme, the moderator enters through one or more inlet nozzles penetrating the calandria shell and flows out through outlet nozzles. Baffles are fixed at the inlet nozzles for proper distribution of moderator in the calandria and to avoid the impact of the jet on the neighbouring calandria tubes. An alternate scheme for moderator inlet has been conceived and engineered in which three lower peripheral lattice locations of the reactor are converted into moderator inlets. This is achieved by moderator sparger channels each containing a 5 m long perforated zircaloy-2 sparger tube rolled to the calandria tube sheets and extended by stainless steel tubular components (inserts) at both ends of a sparger channel. Moderator enters the sparger channel at both ends and flows into the calandria. In the absence of standard codes for design of rolled joints, it was requires to develop these joints based on trials followed by various tests. this paper discusses the details of the rolled joint developed for this purpose, the details of the trials with test results and optimization of rolling parameters for these joints

  6. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  7. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

    1988-01-01

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

  8. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  9. Study on thermal neutron spectra in reactor moderators by time-of-flight method

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-12-01

    Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

  10. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  11. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

  12. The moderator's moderator

    International Nuclear Information System (INIS)

    Williamson, G.K.

    1990-01-01

    A brief account is given of the development of graphite moderators for Magnox and advanced gas cooled reactors. The accident at Windscale in 1957 brought to worldwide attention the importance of irradiation damage in graphite and the consequent storage of Wigner energy. In spite of the Windscale setback, preparations for the civil programme of Magnox reactors went ahead apace. Some of the background to the disastrous Dungeness B tender is presented. In spite of all the difficulties and uncertainties, the graphite in UK reactors has performed well. In all cases, as far as the author is aware, the behaviour of the graphite moderators will not prevent design life being achieved. (author)

  13. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  14. Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok

    2015-01-01

    The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature

  15. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  16. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Guidez, Joel; Jarriand, Paul.

    1975-01-01

    The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

  17. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  18. Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Smith, D.L.

    1988-01-01

    The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs

  19. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  20. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

    Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

  1. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants Refs, figs, tabs

  2. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  3. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  4. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  5. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1993-01-01

    Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

  6. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

  7. Measurement of cold neutron spectra at a model of cryogenic moderator of the IBR-2M reactor

    International Nuclear Information System (INIS)

    Kulikov, S.A.; Chernikov, A.N.; Shabalin, E.P.; Kalinin, I.V.; Morozov, V.M.; Novikov, A.G.; Puchkov, A.V.

    2010-01-01

    The article is dedicated to methods and results of experimental determination of cold neutron spectra from solid mesitylene at neutron moderator temperatures 10-50 K. Experiments were fulfilled at the DIN-2PI spectrometer of the IBR-2 reactor. The main goals of this work were to examine a system of constants for Monte Carlo calculation of cryogenic moderators of the IBR-2M reactor and to determine the temperature dependence of cold neutron intensity from the moderator. A reasonable agreement of experimental and calculation results for mesitylene at 20 K has been obtained. The cold neutron intensity at temperature of moderator 10 K is about 1.8 times higher than at T=50 K

  8. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

    1994-01-01

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  9. Effect of scaling on the thermal hydraulics of the moderator of a CANDU reactor

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2011-01-01

    Three dimensional numerical simulations are conducted on the CANDU Moderator Test Facility (MTF) and the actual size CANDU reactor. Moderator test facility is ¼ scale of the actual reactor. The heat input and other operating conditions are scaled down from the real reactor to the MTF using constant Archimedes number (as considered in MTF experiments performed by Atomic Energy of Canada Ltd.). The heat generations inside both tanks are applied through volumetric heating. In this method, heat is added to the fluid throughout the volume as it occurs in real reactor through fission heat generation and gamma rays from radioactive materials. The temperatures in actual reactor simulation are about 10 deg C greater than in MTF simulations. The separation between high and low temperature zones are more visible in real reactor simulation comparing to MTF simulation. The result indicates that the MTF has better mixing and weaker buoyancy forces comparing to real reactor. The velocity distribution in both cases seems similar with impingement point for inlet jets in both cases at the right hand side of the tank. Although the velocities are considerably higher (about 40%) in the case of real reactor, but as we go toward inner core of the tanks, the velocities are similar and very low. Several points inside the tank are monitored for their temperature and velocity with time. The results for these points show fluctuations in both temperature and velocity inside the tank. The fluctuations frequency seems higher in the case of real reactor while the amplitude of fluctuations is smaller in real reactor in most of the points. Here, in this research we have shown that Archimedes number alone cannot be a good scaling parameter (as used in MTF experiments) and it should be used along with Rayleigh number for scaling purposes. (author)

  10. Present status and future perspective of R and D on lead heavy metal-cooled fast reactors

    International Nuclear Information System (INIS)

    Takahashi, Minoru

    2007-01-01

    Since a lead heavy metal (lead-bismuth eutectic) is chemically inert and has higher boiling point compared to a sodium, a lead heavy metal-cooled fast reactor can be inherently safe and has good nuclear characteristics and is so suitable to a medium-small size of the reactor. R and D on corrosion of a lead heavy metal has been carried out in the world and this issue might be solved to choose specific corrosion resistant alloys for structural materials and fuel cans of a lead heavy metal-cooled reactor. This article reviews present status and future perspective on lead heavy metal-cooled fast reactors. (T. Tanaka)

  11. Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O

    1960-11-15

    The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

  12. Pelletized cold moderator of the IBR-2 reactor: current status and future development

    International Nuclear Information System (INIS)

    Ananiev, V; Beliakov, A; Bulavin, M; Verkhogliadov, A; Kulagin, E; Kulikov, S; Mukhin, K; Shabalin, E; Loktaev, K

    2016-01-01

    Current status and future development of the pelletized cold moderator of the IBR-2 reactor in Neutron Physics Laboratory of JINR are represented. Nowadays cold moderator works for physical experiments and allows conducting experiments in the region of wavelengths more than 4 Å up to 10-13 times faster in comparison with the warm water moderator. Future development of the pelletized cold moderator is aimed at increasing the time of its operation for experiments and is based on three components: creation of a system of continuous charging and discharging of beads, supplementation of various additives, and use of new materials, such as triphenylmethane. (paper)

  13. HTGR Metallic Reactor Internals Core Shell Cutting & Machining Antideformation Technique Study

    International Nuclear Information System (INIS)

    Xing Huiping; Xue Song

    2014-01-01

    The reactor shell assembly of HTGR nuclear power station demonstration project metallic reactor internals is key components of reactor, remains with high-precision large component with large-sized thin-walled straight cylinder-shaped structure, and is the first manufacture in China. As compared with other reactor shell, it has a larger ID (Φ5360mm), a longer length (19000mm), a smaller wall thickness (40mm) and a higher precision requirement. During the process of manufacture, the deformation due to cutting & machining will directly affect the final result of manufacture, the control of structural deformation and cutting deformation shall be throughout total manufacture process of such assembly. To realize the control of entire core shell assembly geometry, the key is to innovate and make breakthroughs on anti-deformation technique and then provide reliable technological foundations for the manufacture of HTGR metallic reactor internals. (author)

  14. Radionuclide trap for liquid metal cooled reactors

    International Nuclear Information System (INIS)

    McGuire, J.C.; Brehm, W.F.

    1978-10-01

    At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

  15. Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2006-01-01

    As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

  16. Materials requirements for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Bennett, J.W.; Horton, K.E.

    1978-01-01

    Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

  17. The liquid hydrogen moderator at the NIST research reactor

    International Nuclear Information System (INIS)

    Williams, Robert E.; Rowe, J. Michael; Kopetka, Paul

    1997-09-01

    In 1995, the NIST research reactor was shut down for a number of modifications, including the replacement of the D 2 O cold neutron source with a liquid hydrogen moderator. When the liquid hydrogen source began operating, the flux of cold neutrons increased by a factor of six over the D 2 O source. The design and operation of the hydrogen source are described, and measurements of its performance are compared with the Monte Carlo simulations used in the design. (auth)

  18. Research on Reduced-Moderation Water Reactor (RMWR)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro

    1999-11-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from 238 U to 239 Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

  19. Research on Reduced-Moderation Water Reactor (RMWR)

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1999-11-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

  20. Addressing safety issues in a hybrid liquid metal reactor

    International Nuclear Information System (INIS)

    Garabedyan, D.; Ehvart, Eh.

    1988-01-01

    Hybrid design of a fast reactor with a liquid-metal coolant, combining the advantages of traditional loop (prevailing in the USA) and integral (prevailing in Europe) arrangements is described. Just as a loop reactor the hybrid one has separate arrangement of the core and the equipment of the primary circuit heat exchange. At the same time similar to the reactor with integral arrangement, the option considered has no complex pipeline system. This reduces sharply the possibility of sodium leakages which cause fires and personnel irradiation

  1. Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

    International Nuclear Information System (INIS)

    Loeb, A.; Sokcic-Kostic, M.; Eisenmann, B.; Prechtl, E.

    2007-01-01

    This paper presents the segmentation of the in 8 meter depth of water and for cutting through super alloyed moderator vessel and of the thermal shield of the MZFR stainless steel up to 130 mm wall thickness. Depending on the research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the MZFR reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m 3 . Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (author)

  2. Liquid metal reactor head designs in the USA - heat and mass transfer considerations

    International Nuclear Information System (INIS)

    Burke, T.M.

    1986-01-01

    Development of liquid metal reactor plants in the United States over the past 30 years has resulted in an evolution of reactor head designs as reflected in the SRE, Hallam, EBR-II and FFTF plants. This evolution has probably been affected to some extent by the fact that, in contrast to most other countries, there is no single organization in the United States which has been responsible for the design of liquid metal reactor plants. The current U.S. LMR design efforts involve two innovative design consortiums (guided by the US Department of Energy) and a joint industry venture on the Large Scale Prototype Breeder. It is therefore somewhat difficult to provide a statement on the philosophy of the reactor head design in the U.S. This paper however briefly describes the existing and proposed U.S. liquid metal reactor head designs and in the process, attempt to provide some insight on the basis for those designs

  3. Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

    International Nuclear Information System (INIS)

    Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

    2001-01-01

    An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

  4. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

  5. The evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Lancaster, D.

    1992-01-01

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

  6. Metal fires and their implications for advanced reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  7. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  8. An automatic regulating control system for a graphite moderated reactor using digital techniques

    International Nuclear Information System (INIS)

    Carvalho Goncalves Junior, J. de.

    1989-01-01

    The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

  9. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  10. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  11. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  12. Simplified analysis of trasients in pool type liquid metal reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1987-01-01

    The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

  13. Status of liquid metal reactor development in the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, Jerry D [Reactor Systems Development and Technology, Office of Nuclear Energy, U.S. Department of Energy (United States); Horton, Kenneth E [Division of International Programs, Office of Nuclear Energy, U.S. Department of Energy (United States)

    1992-07-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  14. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, Jerry D.; Horton, Kenneth E.

    1992-01-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  15. Effect of sulfide concentration on the location of the metal precipitates in inversed fluidized bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villa-Gomez, D., E-mail: d.villagomez@unesco-ihe.org [Core Pollution Prevention and Control, UNESCO-IHE, Institute for Water Education, PO Box 3015, 2601 DA Delft (Netherlands); Ababneh, H.; Papirio, S.; Rousseau, D.P.L.; Lens, P.N.L. [Core Pollution Prevention and Control, UNESCO-IHE, Institute for Water Education, PO Box 3015, 2601 DA Delft (Netherlands)

    2011-08-15

    Highlights: {yields} Sulfide concentration governs the location of metal precipitates in sulfate reducing bioreactors. {yields} High dissolved sulfide induces metal precipitation in the bulk liquid as fines. {yields} Low dissolved sulfide concentrations yield local supersaturation and thus metal precipitation in the biofilm. -- Abstract: The effect of the sulfide concentration on the location of the metal precipitates within sulfate-reducing inversed fluidized bed (IFB) reactors was evaluated. Two mesophilic IFB reactors were operated for over 100 days at the same operational conditions, but with different chemical oxygen demand (COD) to SO{sub 4}{sup 2-} ratio (5 and 1, respectively). After a start up phase, 10 mg/L of Cu, Pb, Cd and Zn each were added to the influent. The sulfide concentration in one IFB reactor reached 648 mg/L, while it reached only 59 mg/L in the other one. In the high sulfide IFB reactor, the precipitated metals were mainly located in the bulk liquid (as fines), whereas in the low sulfide IFB reactor the metal preciptiates were mainly present in the biofilm. The latter can be explained by local supersaturation due to sulfide production in the biofilm. This paper demonstrates that the sulfide concentration needs to be controlled in sulfate reducing IFB reactors to steer the location of the metal precipitates for recovery.

  16. Liquid metal reactor absorber technology

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1990-10-01

    The selection of boron carbide as the reference liquid metal reactor absorber material is supported by results presented for irradiation performance, reactivity worth compatibility, and benign failure consequences. Scram response requirements are met easily with current control rod configurations. The trend in absorber design development is toward larger sized pins with fewer pins per bundle, providing economic savings and improved hydraulic characteristics. Very long-life absorber designs appear to be attainable with the application of vented pin and sodium-bonded concepts. 3 refs., 3 figs

  17. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1992-01-01

    Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

  18. Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor

    Directory of Open Access Journals (Sweden)

    Vorobiev Alexander V.

    2017-01-01

    Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.

  19. Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Downar, T.J.

    1987-01-01

    A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

  20. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-01-01

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF

  1. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  2. Safety design analyses of Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  3. The thorium fuel cycle in water-moderated reactor systems

    International Nuclear Information System (INIS)

    Critoph, E.

    1977-01-01

    Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

  4. Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

    International Nuclear Information System (INIS)

    Hine, R.E.

    1980-09-01

    This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

  5. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.; Gegenheimer, M.

    1984-11-01

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.) [de

  6. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  7. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  8. Fabrication of particulate metal fuel for fast burner reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

    2012-01-01

    U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

  9. Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

    1987-01-01

    A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

  10. A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

    International Nuclear Information System (INIS)

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

  11. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  12. A conceptual design strategy for liquid-metal-wall inertial fusion reactors

    International Nuclear Information System (INIS)

    Monsler, M.J.; Meier, W.R.

    1981-01-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade are reviewed from the perspective of formulating a conceptual design strategy for such chambers. The basis for the design strategy is set by enumerating both the attractive and unattractive features of a LMW chamber. Past concepts are then reviewed to identify conceptual design approaches and physical configurations that enhance the positive aspects and minimize the negative aspects. A detailed description of the engineering considerations is given, including such topics as the selection of a liquid metal, control of radiation damage, selection of structural material, control of tritium breeding and extraction, control of wall stress, and designing for a given rep-rate. Finally, a design strategy is formulated which accomodates the engineering constraints while minimizing the liquid-metal flow rate. (orig.)

  13. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  14. Measurement of neutron disadvantage factor for fuel and moderator in the square reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.; Spiric, V.

    1964-01-01

    Full text: Heterogeneous diffusion treatment for flux distribution was used to define the direction of measurements for obtaining mean neutron flux in the moderator of the reactor cell by single integration. Factor Q for the fuel was determined by using experimental flux distribution in the cell moderator and calculated values for the function X (x;y). Experimental and calculated results are shown as a diagram. All the calculations were done on the ZUSE-Z-23 computer

  15. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

    International Nuclear Information System (INIS)

    Hery, M.; Lecocq, A.

    1983-03-01

    After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

  16. Metal of cavitation erosion of a hydrodynamic reactor

    Science.gov (United States)

    Zakirzakov, A. G.; Brand, A. E.; Petryakov, V. A.; Gordievskaya, E. F.

    2017-02-01

    Cavitation erosion is a major cause of the petroleum equipment hydraulic erosion, which leads to the metal weight loss of the equipment and its breakdown, which can be followed by the full stop of the plant or company work. The probability of the metal weight loss and equipment failure can be reduced by the use of special protective coatings or rivets, made of the sacrificial metals, the use of which significantly increases the service life and the production equipment reliability. The article investigates the cavitation erosion effect, occurred under the condition of the advanced hydrodynamic cavitation on the hydrodynamic cavitation reactor. This article presents the results of the experiments and recommendations for increasing the operational resource.

  17. Metallic materials corrosion problems in molten salt reactors

    International Nuclear Information System (INIS)

    Chauvin, G.; Dixmier, J.; Jarny, P.

    1977-01-01

    The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 700 0 C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm -2 at 700 0 C [fr

  18. Summary report of the 7th reduced-moderation water reactor workshop

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

    2005-08-01

    As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

  19. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

    1998-03-01

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  20. Progress in design study on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Shirakawa, Toshihisa; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takeda, Renzo [Hitachi Ltd., Tokyo (Japan); Yokoyama, Tsugio [Toshiba Corp., Kawasaki, Kanagawa (Japan); Hibi, Koki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Wada, Shigeyuki [Japan Atomic Power Co., Tokyo (Japan)

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPC) in 1998, under technical cooperation with three Japanese reactor vendors. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight-lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR type core with high void fraction and super-flat core, a long operation cycle BWR type core using void tube assembly, a high conversion BWR type core without blankets, a high conversion PWR type core using heavy water as a coolant, and a PWR type core for plutonium multi-recycle using seed-blanket type fuel assemblies. Detailed feasibility studies for the RMWR have been continued on core design study. The present report summarizes the recent progress in the design study for the RMWR. (author)

  1. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  2. Fast reactors with axial arrangement of oxide and metal fuels in the core

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

    1980-01-01

    Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

  3. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    International Nuclear Information System (INIS)

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  4. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  5. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Merk, B.; Weiß, F.P.

    2011-01-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  6. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Weiß, F.P., E-mail: b.merk@fzd.de [Forschungszentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden (germany)

    2011-07-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  7. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Solomon, K.A.

    1979-07-01

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  8. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  9. Evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The project concentrated on model development. Reactor physics modeling involved establishing accurate models with PC versions of COMBINE and VENTURE. Fuel performance analysis will start with METAL- LIFE. In order to justify the change of fuel to metal alloy, large benefits will have to be found; the cost benefit reported is not sufficient. The fuel pin will be annular and contact the clad; the clad thickness will force the fuel to grow toward the central hole. This report reports: design improvements, neutronic model development, COBRA modifications, reactor kinetics model development, RELAP code, and fuel performance

  10. Cirus reactor: a milestone in Indian Atomic Energy Programme

    International Nuclear Information System (INIS)

    Ranjan, Rakesh; Karhadkar, C.G.; Bhattacharya, S.

    2017-01-01

    Cirus, a 40 MW_t_h, high flux, thermal neutron research reactor achieved first criticality on 10"t"h July 1960. It had vertical core, natural metallic Uranium rods in Aluminium clad as fuel, demineralised light water as coolant, heavy water as moderator, Helium as cover gas and graphite as reflector. A low-pressure containment was provided for the reactor and some of the important associated reactor systems. Reactor start-up and power regulation was effected by controlled adjustment of moderator level in the reactor vessel. Boron carbide rods were used as primary shut down devices. Dumping of heavy water from core worked as secondary shut down device. Seawater was used as secondary coolant for removal of the fission heat of the reactor. Initial operation of Cirus was marred by several difficulties, primarily arising out of water chemistry in primary cooling water system. It took almost 3 years to systematically resolve these problems and achieve stable operation of reactor. Cirus could be operated at its rated power by the year 1963

  11. Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

  12. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  13. On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@fzd.de [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany); Fridman, Emil; Weiss, Frank-Peter [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany)

    2011-05-15

    Research highlights: > Using a moderation layer can reduce the sodium void effect in a SFR. > Inserting the moderation layer improves the Doppler effect significantly. > The uniform layer distribution avoids effects on power and burnup distribution. > Hydride containing material like uranium-zirconium hydride is most efficient. - Abstract: This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B{sub 4}C or uranium-zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.

  14. Minor Actinides Burnup Enhancement in the European Sodium Fast Reactor through Moderator Material Addition

    International Nuclear Information System (INIS)

    Ramos, R.L.; Buiron, L.

    2013-01-01

    Conclusions: • ZrH 2 was the best moderator material, followed by MgO and MgAl 2 O 4 ; • When the number of moderator pins is increased: – the percentage of minor actinides consumed increases; – the total mass consumed of minor actinides decreases; – the decay heat generated decreases; – the neutron flux in the reactor varies very little. Perspectives: • For future studies it would be possible to evaluate the use of other materials with resonances in the scattering cross section in the fast range that would improve the results obtained with Mg. • It would be necessary to consider how to add moderator material without changing the initial mass of minor actinides. E.g., adding the moderator at the periphery of the minor actinide elements

  15. Bioleaching of metals from electronic scrap by moderately thermophilic acidophilic bacteria

    NARCIS (Netherlands)

    Ilyas, Sadia; Anwar, Munir A.; Niazi, Shahida B.; Ghauri, M. Afzal

    The present work was aimed at studying the bioleachability of metals from electronic scrap by the selected moderately thermophilic strains of acidophilic chemolithotrophic and acidophilic heterotrophic bacteria. These included Sulfobacillus thermosulfidooxidans and an unidentified acidophilic

  16. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  17. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  18. Development of a silicon calorimeter for dosimetry applications in a water-moderated reactor

    International Nuclear Information System (INIS)

    DePriest, Kendall Russell; King, Donald Bryan; Naranjo, Gerald E.; Luker, Spencer Michael; Keltner, Ned R.; Suo-Anttila, Ahti Jorma; Griffin, Patrick Joseph

    2005-01-01

    High fidelity active dosimetry in the mixed neutron/gamma field of a research reactor is a very complex issue. For passive dosimetry applications, the use of activation foils addresses the neutron environment while the use of low neutron response CaF 2 :Mn thermoluminescent dosimeters (TLDs) addresses the gamma environment. While radiation-hardened diamond photoconducting detectors (PCD) have been developed that provide a very precise fast response (picosecond) dosimeter and can provide a time-dependent profile for the radiation environment, the mixed field response of the PCD is still uncertain and this interferes with the calibration of the PCD response. In order to address the research reactor experimenter's need for a dosimeter that reports silicon dose and dose rate at a test location during a pulsed reactor operation, a silicon calorimeter has been developed. This dosimeter can be used by itself to provide a dose in rad(Si) up to a point in a reactor pulsed operation, or, in conjunction with the diamond PCD, to provide a dose rate. This paper reports on the development, testing, and validation of this silicon calorimeter for applications in water-moderated research reactors.

  19. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  20. Design of a graphite-moderated {sup 241}Am-Li neutron field to simulate reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, N., E-mail: tsujimura.norio@jaea.go.j [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Yoshida, T. [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan)

    2010-12-15

    A neutron calibration field using {sup 241}Am-Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the {sup 241}Am-Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.

  1. Multidimensional space-time kinetics of a heavy water moderated nuclear reactor

    International Nuclear Information System (INIS)

    Winn, W.G.; Baumann, N.P.; Jewell, C.E.

    1980-01-01

    Diffusion theory analysis of a series of multidimensional space-time experiments is appraised in terms of the final experiment of the series. In particular, TRIMHX diffusion calculations were examined for an experiment involving free-fall insertion of a 235 U-bearing rod into a heavy water moderated reactor with a large reflector. The experimental transient flux-tilts were accurately reproduced after cross section adjustments forced agreement between static diffusion calculations and static reactor measurements. The time-dependent features were particularly well modeled, and the bulk of the small discrepancies in space-dependent features should be removable by more refined cross-section adjustments. This experiment concludes a series of space-time experiments that span a wide range of delayed neutron holdback effects. TRIMHX calculations of these experiments demonstrate the accuracy of the modeling employed in the code

  2. Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

    1989-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

  3. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  4. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  5. Metal-fuel modeling for inherently safe reactor designs

    International Nuclear Information System (INIS)

    Miles, K.J. Jr.

    1987-01-01

    Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

  6. Conceptual study of a complementary scram system for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vanmaercke, S.; Van den Eynde, G.; Tijskens, E.; Bartosiewicz, Y.

    2009-01-01

    GEN-IV reactors promise higher safety and reliability as one of the major improvements over previous generations of reactors. To achieve that, all GEN-IV reactor concepts require two completely independent shutdown systems that rely on different operating principles. For liquid metal cooled reactors the first system is an absorber-rod based solution. The second system that by requirement should rely on another principle, is however quite a challenge to design. The second system used in current PWR reactors is to dissolve a neutron absorber, boric acid, into the primary coolant. This method cannot be used in liquid metal cooled reactors because of the high cost of cleaning the coolant. In this paper an overview of the existing literature on scram systems is given, each with their advantages and limitations. A promising new concept is also presented. This concept leads to a totally passive self activating device using small absorbing particles that flow into a dedicated channel to shutdown the reactor. The system consists of tubes filled with particles of an absorber material. During normal operation, these particles are kept above the active core by means of a metallic seal. In case of an accident, the system is activated by the temperature increase in the coolant. This leads to melting of the metal seal. The ongoing work conducted at SCK·CEN and UCL/TERM aims at assessing the reliability of this new concept both experimentally and numerically. This study is multidisciplinary as neutronic and thermal hydraulics issues are tackled. Most challenging is however the thermal hydraulics related to understanding and predicting the liberation and flow of the absorber particles during a shutdown. Simple experiments are envisaged to compare to numerical simulations using the Discrete Element Method for simulating the particles. In a later stage this will be coupled with Smoothed Particles Hydrodynamics for simulating the melting of the seal. Some preliminary experimental and

  7. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    International Nuclear Information System (INIS)

    Fukaya, Y.; Okubo, T.; Uchikawa, S.

    2008-01-01

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

  8. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2008-07-15

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

  9. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  10. Improvements in gas supply systems for heavy-water moderated reactors

    International Nuclear Information System (INIS)

    Aubert, G.; Hassig, J.M.; Laurent, N.; Thomas, B.

    1964-01-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [fr

  11. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Hahn, D.

    2001-01-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  12. FBR type reactor

    International Nuclear Information System (INIS)

    Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

    1996-01-01

    In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

  13. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  14. Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

    International Nuclear Information System (INIS)

    Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

    1996-01-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

  15. Natural-circulation flow pattern during the gamma-heating phase of an LBLOCA in a heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Rodriguez, S.B.; Unal, C.; Pasamehmetoglu, K.O.; Motley, F.E.

    1992-01-01

    In a postulated large-break loss-of-coolant accident (LBLOCA), the core of the reactor is uncovered quickly as the liquid that drains out of the tank is replaced by air. During the LBLOCA, the reactor is scrammed. the moderator tank is drained, and fuel and control rod tubes are cooled internally by forced convection via the emergency cooling system (ECS) water. However, the safety rods, reflector assemblies, tank wall, and instrument rods continue to heat up as a result of gamma deposition. These components are primarily cooled by natural/mixed convection and radiation heat transfer. In this paper, the thermal-hydraulic analysis of a reactor moderator tank exposed to air during an LBLOCA is discussed. The analysis was performed using a special version of the Transient Reactor Analysis Code (TRAC). TRAC input and code modifications considered the appropriate modeling of ECS cooling, thermal radiation heat transfer, and natural convection. The major objective of the model was to calculate the limiting component temperature (that establishes the maximum operating power) as a result of gamma heating. In addition, the nature of the moderator tank air-circulation pattern and its effects on the limiting temperature under various conditions were analyzed. None of the components were found to exceed their structural limits when the pre-scram power level was 50% of historical power

  16. Some Examples of Accident Analyses for RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  18. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  19. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, A. dos; Nascimento, J.A. do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  20. Methanol degradation in granular sludge reactors at sub-optimal metal concentrations: role of iron, nickel and cobalt

    NARCIS (Netherlands)

    Zandvoort, M.H.; Geerts, R.; Lettinga, G.; Lens, P.N.L.

    2003-01-01

    The effect of sub-optimal trace metal concentrations on the conversion of methanol in an upflow anaerobic sludge bed (UASB) reactor was investigated by studying the effect of decreased influent trace metal concentrations on the reactor efficiency, methanol conversion route and sludge

  1. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  2. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  3. Metal removal efficiency and ecotoxicological assessment of field-scale passive treatment biochemical reactors

    Science.gov (United States)

    Anaerobic biochemical reactors (BCRs) are useful for removing metals from mining-impacted water (MIW) at remote sites. Removal processes include sorption and precipitation of metal sulfides, carbonates and hydroxides. A question of interest is whether BCRs remove aquatic toxicit...

  4. Water storage of liquid-metal fast-breeder-reactor fuel

    International Nuclear Information System (INIS)

    Meacham, S.A.

    1982-01-01

    The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry

  5. The feasibility study of using deuterated gadolinium nitrate for moderator-poisoned shutdown and excess reactivity control in CANDU reactors

    International Nuclear Information System (INIS)

    Li, J.; Everatt, A.

    2006-01-01

    Gadolinium nitrate is used in CANDU stations as moderator poison for reactor shutdowns and excess reactivity control. The use of the light-water hydrate introduces significant quantities of light water into the moderator system, which must be removed from the moderator by periodically upgrading the moderator (isotopic maintenance). The benefit of using a deuterated gadolinium nitrate would be a higher moderator isotopic and/or a lesser isotopic maintenance requirement. This study evaluated the economics of using deuterated gadolinium nitrate, as opposed to the light-water hydrate, for moderator-poisoned shutdowns and excess reactivity control in CANDU-6 reactors. Normal gadolinium nitrate (i.e., the light-water hydrate) is available from suppliers at ∼125 $/kg. Supplier quotes for deuterated gadolinium nitrate ranged from 1900 to 4000 $/kg. To examine the possibility of producing deuterated gadolinium nitrate in-house at a lower cost than commercially available, a three-stage dissolution/evaporation manufacturing process was conceived and costed. Depending on the assumed demand for the product (i.e., the number of reactors adopting the use of the product) and the capital recovery period, the estimated unit cost for the dissolution/evaporation process ranged from 730 to 2500 $/kg. The determination of economic benefit from using deuterated gadolinium nitrate in existing CANDU stations was based on the cost savings resulting from a higher fuel burn-up (i.e., the higher moderator isotopic would give a higher fuel burn-up). The net benefit of using deuterated gadolinium nitrate for most CANDU stations was determined to be marginal (i.e., <20 k$/a). Only for those CANDU stations where the moderator isotopic was relatively low (e.g., 99.85 wt%) was there a potential significant benefit (20-100 k$/a). However, if the reason for the low moderator isotopic is a relatively high moderator light-water ingress rate from sources other than the use of the light-water hydrate

  6. Conceptual safety design analysis of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Suk, S. D.; Park, C. K.

    1999-01-01

    The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

  7. Comparison of oxide- and metal-core behavior during CRBRP [Clinch River Breeder Reactor Plant] station blackout

    International Nuclear Information System (INIS)

    Polkinghorne, S.T.; Atkinson, S.A.

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs

  8. Synthesis of colloidal metal nanocrystals in droplet reactors: the pros and cons of interfacial adsorption.

    Science.gov (United States)

    Zhang, Lei; Wang, Yi; Tong, Limin; Xia, Younan

    2014-07-09

    Droplet reactors have received considerable attention in recent years as an alternative route to the synthesis and potentially high-volume production of colloidal metal nanocrystals. Interfacial adsorption will immediately become an important issue to address when one seeks to translate a nanocrystal synthesis from batch reactors to droplet reactors due to the involvement of higher surface-to-volume ratios for the droplets and the fact that nanocrystals tend to be concentrated at the water-oil interface. Here we report a systematic study to compare the pros and cons of interfacial adsorption of metal nanocrystals during their synthesis in droplet reactors. On the one hand, interfacial adsorption can be used to generate nanocrystals with asymmetric shapes or structures, including one-sixth-truncated Ag octahedra and Au-Ag nanocups. On the other hand, interfacial adsorption has to be mitigated to obtain nanocrystals with uniform sizes and controlled shapes. We confirmed that Triton X-100, a nonionic surfactant, could effectively alleviate interfacial adsorption while imposing no impact on the capping agent typically needed for a shape-controlled synthesis. With the introduction of a proper surfactant, droplet reactors offer an attractive platform for the continuous production of colloidal metal nanocrystals.

  9. Noise analysis method for monitoring the moderator temperature coefficient of pressurized water reactors: Neural network calibration

    International Nuclear Information System (INIS)

    Thomas, J.R. Jr.; Adams, J.T.

    1994-01-01

    A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base

  10. 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

  11. Buckling and reaction rate experiments in plutonium/uranium metal fuelled, graphite moderated lattices at temperatures up to 400 deg. C. Part I: Experimental techniques and results

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Clarke, W G; Gibson, M; Hobday, R; Hunt, C; Marshall, J; Puckett, B J; Symons, C R; Wass, T [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-07-15

    This report presents experimental measurements of bucklings, flux fine structure and fission rate distributions in graphite moderated lattices fuelled with plutonium/uranium metal at temperatures up to 400 deg. C in the sub-critical assemblies SCORPIO I and SCORPIO II. The experimental techniques employed are described in some detail. The accuracy of the experimental measurements appears to be adequate for testing methods of calculation being developed for the calculation of reactivity and temperature coefficient of reactivity for power reactors containing plutonium and uranium. (author) 26 refs, 17 tabs, 17 figs

  12. Review of accident analyses of RB experimental reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2003-01-01

    The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

  13. Review of accident analyses of RB experimental reactor

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2003-01-01

    Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

  14. Fuel transfer manipulator for liquid metal nuclear reactors

    International Nuclear Information System (INIS)

    Sturges, R.H.

    1983-01-01

    A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

  15. Liquid metal pump for nuclear reactors

    International Nuclear Information System (INIS)

    Allen, H.G.; Maloney, J.R.

    1975-01-01

    A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank

  16. A possible approach to 14MeV neutron moderation: A preliminary study case.

    Science.gov (United States)

    Flammini, D; Pilotti, R; Pietropaolo, A

    2017-07-01

    Deuterium-Tritium (D-T) interactions produce almost monochromatic neutrons with about 14MeV energy. These neutrons are used in benchmark experiments as well as for neutron cross sections assessment in fusion reactors technology. The possibility to moderate 14MeV neutrons for purposes beyond fusion is worth to be studied in relation to projects of intense D-T sources. In this preliminary study, carried out using the MCNP Monte Carlo code, the moderation of 14MeV neutrons is approached foreseeing the use of combination of metallic materials as pre-moderator and reflectors coupled to standard water moderators. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Bioleaching of multiple metals from contaminated sediment by moderate thermophiles.

    Science.gov (United States)

    Gan, Min; Jie, Shiqi; Li, Mingming; Zhu, Jianyu; Liu, Xinxing

    2015-08-15

    A moderately thermophilic consortium was applied in bioleaching multiple metals from contaminated sediment. The consortium got higher acidification and metals soubilization efficiency than that of the pure strains. The synergistic effect of the thermophilic consortium accelerated substrates utilization. The utilization of substrate started with sulfur in the early stage, and then the pH declined, giving rise to making use of the pyrite. Community dynamic showed that A. caldus was the predominant bacteria during the whole bioleaching process while the abundance of S. thermotolerans increased together with pyrite utilization. Solubilization efficiency of Zn, Cu, Mn and Cd reached 98%, 94%, 95%, and 89% respectively, while As, Hg, Pb was only 45%, 34%, 22%. Logistic model was used to simulate the bioleaching process, whose fitting degree was higher than 90%. Correlation analysis revealed that metal leaching was mainly an acid solubilization process. Fraction analysis revealed that metals decreased in mobility and bioavailability. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

  19. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  20. The US Advanced Liquid-Metal Reactor Program

    International Nuclear Information System (INIS)

    Brolin, E.C.

    1992-01-01

    Based on National Energy Strategy projections, utilities will be required to substantially increase electric generating capacity over the next 40 yr to meet economic growth requirements and replace retiring capacity. Although aggressive conservation measures can save up to 85 GW(electric), ∼195 GW(electric) of additional generating capcity will still be needed by 2010. Assuming startup of new plants around 2000, US Department of Energy (DOE) analyses show that nuclear power can contribute 195 GW(electric) of capacity by 2030, or ∼20% of total electric generation. The DOE is involved in a number of strategies designed to revitalize the nuclear power industry and enable it to meet this projected need for additional capacity. Among these is an integrated overall strategy for advanced reactor development and high-level waste management. A high priority in pursuit of this strategy is the Advanced Liquid-Metal Reactor (ALMR) Program

  1. Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

    International Nuclear Information System (INIS)

    Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

    1989-01-01

    The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

  2. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  3. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  4. Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

    Science.gov (United States)

    Hvasta, M. G.; Kolemen, E.; Fisher, A. E.; Ji, H.

    2018-01-01

    Plasma-facing components (PFC’s) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC’s, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC’s can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metal that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. These results show the promise of electromagnetic control for LM-PFC’s and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.

  5. Safety aspects of the US advanced LMR [liquid metal reactor] design

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.

    1989-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs

  6. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  7. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  9. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  10. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  11. A study concerning tritium concentration evolution in the moderator of a CANDU reactor connected on-line to a detritiation facility

    International Nuclear Information System (INIS)

    Bidica, Nicolae; Bornea, Anisia

    2005-01-01

    The present work is a theoretical study on the tritium concentration evolution in the CANDU reactor moderator connected on-line with a detritiation facility. This study is based on a calculation model which takes into account the evolution curve of the tritium concentration in the absence of detritiation process in both the moderator and SPTC of the Unit 1 CANDU reactor at Cernavoda NPP. This study leads to determination of the tritium concentration evolution in the moderator in the presence of the detritiation process for both a range of intake flows and initial concentration. Also, the intake flow change will be analyzed for a detritiation facility as a function of tritium initial concentration existing in the moderator in the case of a survey of the detritiation over a given period of time. The conclusions of this study were the following: - an optimum of the detritiation factor can be determined; - detritiation starts at a lower value for the tritium concentration in moderator which reduces the strain upon the detritiation facility and therefore the costs of its building, maintenance and operation. (authors)

  12. Assessment of corrosion and fatigue damage to light water reactor metal containments

    International Nuclear Information System (INIS)

    Sinha, U.P.; Shah, V.N.; Smith, S.K.

    1991-01-01

    This paper presents a generic procedure for estimating aging damage, evaluating structural integrity, and identifying mitigation activities for safe operation of boiling water reactor (BWR) Mark I metal containments and ice-condenser type pressurized water reactor (PWR) cylindrical metal containments. The mechanisms of concern that can cause aging damage to these two types of containments are corrosion and fatigue. Assessment of fatigue damage to bellows is also described. Assessment of corrosion and fatigue damage described in this paper include: containment design features that are relevant to aging assessment, several corrosion and fatigue mechanisms, inspection of corrosion and fatigue damage, and mitigation of damage caused by these mechanisms. In addition, synergistic interaction between corrosion and fatigue is considered. Possible actions for mitigating aging include enhanced inspection methods, maintenance activities based on operating experience, and supplementary surveillance programs. Field experience related to aging of metal containments is reviewed. Finally, conclusions and recommendations are presented

  13. Validation of moderator-level reactivity coefficient using station data

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M.; Martchouk, I., E-mail: mohamed.younis@amecfw.com, E-mail: iouri.martchouk@amecfw.com [Amec Foster Wheeler, Toronto, ON (Canada); Buchan, P.D., E-mail: david.buchan@opg.com [Ontario Power Generation, Pickering, ON (Canada)

    2015-07-01

    The reactivity effect due to variations in the moderator level has been recognized as a reactor physics phenomenon of importance during normal operation and accident analysis. The moderator-level reactivity coefficient is an important parameter in safety analysis of CANDU reactors, e.g., during Loss of Moderator Heat Sink as well as in the simulation of Reactor Regulating System action in CANDU reactors that use moderator level for reactivity control. This paper presents the results of the validation exercise of the reactor-physics toolset using the measurements performed in Pickering Unit 4 in 2003. The capability of the code suite of predicting moderator-level reactivity effect was tested by comparing measured and predicted reactor-physics parameters. (author)

  14. 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

  15. Reactor prospects of muon-catalyzed fusion of deuterium and tritium concentrated in transition metals

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1989-01-01

    It is conjectured that the number of fusion events catalyzed by a single muon is orders of magnitude greater for deuterium and tritium concentrated in a transition metal than in gaseous form and that the recent observation of 2.5-MeV neutrons from a D 2 O electrolytic cell with palladium and titanium cathodes can thereby be interpreted in terms of cosmic muon-catalyzed deuterium-deuterium fusion. This suggests a new fusion reactor reactor consisting of deuterium and tritium concentrated in transition metal fuel elements in a fusion core that surrounds an accelerator-produced muon source. The feasibility of net energy production in such a reactor is established in terms of requirements on the number of fusion events catalyzed per muon. The technological implications for a power reactor based on this concept are examined. The potential of such a concept as a neutron source for materials testing and tritium and plutonium production is briefly discussed

  16. Passive safety and the advanced liquid metal reactors

    International Nuclear Information System (INIS)

    Hill, D.J.; Pedersen, D.R.; Marchaterre, J.F.

    1988-01-01

    Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental programs in this area

  17. Effect of hydraulic retention time on metal precipitation in sulfate reducing inverse fluidized bed reactors

    KAUST Repository

    Villa-Gómez, Denys Kristalia

    2014-02-13

    BACKGROUND: Metal sulfide recovery in sulfate reducing bioreactors is a challenge due to the formation of small precipitates with poor settling properties. The size of the metal sulfide precipitates with the change in operational parameters such as pH, sulfide concentration and reactor configuration has been previously studied. The effect of the hydraulic retention time (HRT) on the metal precipitate characteristics such as particle size for settling has not yet been addressed. RESULTS: The change in size of the metal (Cu, Zn, Pb and Cd) sulfide precipitates as a function of the HRT was studied in two sulfate reducing inversed fluidized bed (IFB) reactors operating at different chemical oxygen demand concentrations to produce high and low sulfide concentrations. The decrease of the HRT from 24 to 9h in both IFB reactors affected the contact time of the precipitates formed, thus making differences in aggregation and particle growth regardless of the differences in sulfide concentration. Further HRT decrease to 4.5h affected the sulfate reducing activity for sulfide production and hence, the supersaturation level and solid phase speciation. Metal sulfide precipitates affected the sulfate reducing activity and community in the biofilm, probably because of the stronger local supersaturation causing metal sulfides accumulation in the biofilm. CONCLUSIONS: This study shows that the HRT is an important factor determining the size and thus the settling rate of the metal sulfides formed in bioreactors.

  18. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

  19. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

  20. Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

    International Nuclear Information System (INIS)

    Anand, G.; Christensen, R.N.

    1990-01-01

    The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

  1. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  2. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    International Nuclear Information System (INIS)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2016-01-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  3. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  4. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  5. Graphite moderated 252Cf source

    International Nuclear Information System (INIS)

    Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

    2014-08-01

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  6. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  7. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  8. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  9. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  10. Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

    1987-01-01

    This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

  11. Contribution to the use of a solid moderator gas reactor, for naval propulsion

    International Nuclear Information System (INIS)

    Pheline, J.; Gautier, A.

    1960-01-01

    In this contribution, the authors discuss works performed in France for the development of nuclear propulsion in merchant ships, notably for an oil tanker of 50.000 tons with 17 knot speed, i.e. a 20.000 Hp engine with an energy produced by a 60 MW gas reactor with a solid moderator and comprising 400 channels loaded with uranium oxide enriched ay 2.8 per cent and sheathed with a refractory alloy. The authors discuss the possible materials for the moderator, the heat transfer medium, the sheath, the fuel and the structures, and report technological studies (mechanical tests, irradiation tests) performed to investigate material properties and their behaviour in operation conditions. They report tests performed to investigate core structure characteristics with respect to neutrons. They finally briefly present a prototype

  12. Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

    1978-07-01

    The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

  13. Trade study for kWe class space reactors

    Science.gov (United States)

    Bost, Donald S.

    Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

  14. Apparatus for sealing a rotatable shield plug in a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1980-01-01

    An apparatus for sealing a rotatable shield plug in a nuclear reactor having liquid metal coolant is described. The apparatus includes a dip -ring seal adapted to provide a fluid barrier between the liquid metal and the atmosphere and to permit rotation of the shield plug. The apparatus also includes a static seal for the rotatable shield plug located between the dip-ring seal and the liquid metal. The static seal isolates the dip-ring seal from the liquid metal vapor during operation at power and can be disengaged for rotation of the shield plug

  15. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  16. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  17. Adaptability of Brayton cycle conversion systems to fast, epithermal and thermal spectrum space nuclear reactors

    International Nuclear Information System (INIS)

    Tilliette, Z.P.

    1988-01-01

    The two French Government Agencies C.N.E.S. (Centre National d'Etudes Spatiales) and C.E.A. (Commissariat a l'Energie Atomique) are carrying out joint preliminary studies on space nuclear power systems for future ARIANE 5 launch vehicle applications. The Brayton cycle is the reference conversion system, whether the heat source is a liquid metal-cooled (NaK, Na or Li) reactor or a gas-cooled direct cycle concept. The search for an adequate utilization of this energy conversion means has prompted additional evaluations featuring the definition of satisfactory cycle conditions for these various kinds of reactor concepts. In addition to firstly studied fast and epithermal spectrum ones, thermal spectrum reactors can offer an opportunity of bringing out some distinctive features of the Brayton cycle, in particular for the temperature conditioning of the efficient metal hydrides (ZrH, Li/sub 7/H) moderators. One of the purposes of the paper is to confirm the potential of long lifetime ZrH moderated reactors associated with a gas cycle and to assess the thermodynamical consequences for both Nak(Na)-cooled or gas-cooled nuclear heat sources. This investigation is complemented by the definition of appropriate reactor arrangements which could be presented on a further occasion

  18. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

    International Nuclear Information System (INIS)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

  19. Systems analysis of a 100-MWe modular liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Morris, E.E.; Rhow, S.K.; Switick, D.M.

    1985-01-01

    The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

  20. Diversion analysis and safeguards measures for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1981-10-01

    The general objective of the study is to perform a diversion analysis and an assessment of the available safeguards methods and systems for verifying inventory and flow of nuclear material in accessible and inaccessible areas of liquid-metal fast breeder reactor, LMFBR, systems. The study focuses primarily on the assembly-handling operations, assembly storage facilities, and reactor operations facilities relating to existing and/or near-term planned experimental, demonstration and prototypal reactor plants. The safeguards systems and methods presented are considered to be feasible for development and for implementation within the resource limitation of the IAEA and are considered to be consistent with the objectives, requirements, and constraints of the IAEA as outlined in the IAEA documents INFCIRC/153 and INFCIRC/66-Rev-2

  1. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

    2004-01-01

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  2. Evaluating advanced LMR [liquid metal reactor] reactivity feedbacks using SSC

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.; Cheng, H.S.

    1988-01-01

    Analyses of the PRISM and SAFR Liquid Metal Reactors with SSC are discussed from a safety and licensing perspective. The PRISM and SAFR reactors with metal fuel are designed for inherent shutdown responses to loss-of-flow and loss-of-heat-sink events. The demonstration of this technology was performed by EBR-II during experiments in April 1986 by ANL (Planchon, et al.). Response to postulated TOPs (control rod withdrawal) are made acceptable largely by reducing reactivity swings, and therefore minimizing the size of possible ractivity insertions. Analyses by DOE and the contractors GE, RI, and ANL take credit for several reactivity feedback mechanisms during transient calculations. These feedbacks include Doppler, sodium density, and thermal expansion of the grid plates, the load pads, the fuel (axial) and the control rod which are now factored into the BNL SSC analyses. The bowing feedback mechanism is not presently modeled in the SSC due to its complexity and subsequent large uncertainty. The analysis is conservative by not taking credit for this negative feedback mechanism. Comparisons of BNL predictions with DOE contractors are provided

  3. Investigation of noble metal deposition behaviour in boiling water reactors. The NORA project

    International Nuclear Information System (INIS)

    Ritter, Stefan; Karastoyanov, Vasil; Abolhassani-Dadras, Sousan; Guenther-Leopold, Ines; Kivel, Niko

    2010-01-01

    NobleChem trademark is a technology developed by General Electric to reduce stress corrosion cracking (SCC) in reactor internals and recirculation pipes of boiling water reactors (BWRs) while preventing the negative side effects of classic hydrogen water chemistry. Noble metals (Pt, Rh) acting as electrocatalysts for the recombination of O 2 and H 2 O 2 with H 2 to H 2 O and thus reducing the corrosion potential more efficiently are injected into the feedwater during reactor shutdown (classic method) or on-line during power operation. They are claimed to deposit as very fine metallic particles on all water-wetted surfaces, including the most critical regions inside existing cracks, and to stay electrocatalytic over long periods of time. The effectiveness of this technology in plants still remains to be demonstrated. Based on highly credible laboratory experiments down to the sub-μg . kg -1 Pt concentration range, SCC mitigation may be expected, provided that a stoichiometric excess of H 2 and a sufficient surface coverage with very fine Pt particles exist simultaneously at the critical locations [1]. Very little is known about the deposition and (re-)distribution behaviour of the Pt in the reactor. For the validation of this technique the research project NORA (noble metal deposition behaviour in BWRs) has been started at the Paul Scherrer Institute (PSI) with two main objectives: (i) to gain phenomenological insights and a better basic understanding of the Pt distribution and deposition behaviour in BWRs; (ii) to develop and qualify a non-destructive technique to characterise the size and distribution of the Pt particles and the local concentration of Pt on reactor components. This paper presents the objectives of the project, the planned work and a brief description of the status of the project. (orig.)

  4. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

    2015-01-01

    Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

  5. Fuel rod for liquid metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vinz, P.

    1976-01-01

    In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

  6. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  7. Moderator circulation in CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Hussein, M.A.

    1989-01-01

    A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

  8. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  9. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  10. Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

    Science.gov (United States)

    Kennedy, Daniel; Jaworski, Michael

    2014-10-01

    Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

  11. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  12. Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

  13. Homogenous reactor, elaborations, not released up to end

    International Nuclear Information System (INIS)

    Takibayev, Zh.S.

    2002-01-01

    Nowadays the nuclear power uses mainly water moderated reactors, where water or heavy water works as neutron inhibitor or coolant, and fuel solid state is situated in reactor core discretely as fuel element packed in fuel assembly. Such fuel composition in solid state reactors leads to rise in price of reactor itself and, of course, many other inconveniences. Firstly, burning out depth is limited; secondary, agents absorbed neutrons are accumulated in fission products, i. e. it leads to poisoning slag derive and thirdly, there are too many outside agents in reactor core in the form of fuel elements and different constructional materials. It worsens neutron balance of reactor. There are many other inconveniences. Specialists understand this problem. They are looking for escaping of difficulty proposing to begin a wide-ranging design, for example, of a new generation of homogeneous reactor especially with salt liquid, liquid metal fuel. But this problem nowadays can not be nearly decided. It is clear enough that within at least 50-100 years the existing monopoly will not change its attitude to use of new elaboration, for example, reactor with salt liquid fuel unless a sharp necessity of opening up not only 1-2 % of uranium in the case of reactors on thermal neutrons or nearby 10-20 % for fast reactors as nowadays but effective use of all potential of nuclear fission energy contained in natural uranium and thorium resources will be realized. In the report the scheme of nuclear reactor with liquid metal or salt liquid is shown. Such approach can be in future one of possible variants of problem solution in effective opening up of all uranium-plutonium energy resource of our planet. The scheme shows only possible allocations of the container and the pipeline. Their proportioning is one of main problems of future elaborations. A mutual allocation of the container and pipelines was carried out in such way, that demand to the last ones where less than to the container

  14. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  15. Induction apparatus monitoring structural strains in liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Dean, S.A.; Evans, R.A.

    1981-01-01

    An improved method of monitoring induced torsional and linear strains in the internal structures of liquid metal cooled nuclear reactors is described. An electrical induction apparatus indicates the variation of magnetic coupling caused by a ferromagnetic member of the apparatus being subjected to such strains. (U.K.)

  16. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  17. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  18. Development of a noise-based method for the determination of the moderator temperature coefficient of reactivity (MTC) in pressurized water reactors (PWRs)

    International Nuclear Information System (INIS)

    Demaziere, C.

    2002-01-01

    The Moderator Temperature Coefficient of reactivity (MTC) is an important safety parameter of Pressurized Water Reactors (PWRs). In most countries, the so-called at-power MTC has to be measured a few months before the reactor outage, in order to determine if the MTC will not become too negative. Usually, the at-power MTC is determined by inducing a change in the moderator temperature, which has to be compensated for by other means, such as a change in the boron concentration. An MTC measurement using the boron dilution method is analysed in this thesis. It is demonstrated that the uncertainty of such a measurement technique is so large, that the measured MTC could become more negative than what the Technical Specifications allow. Furthermore, this technique incurs a disturbance of the plant operation. For this reason, another technique relying on noise analysis was proposed a few years ago. In this technique, the MTC is inferred from the neutron noise measured inside the core and the moderator temperature noise measured at the core-exit, in the same or in a neighbouring fuel assembly. This technique does not require any perturbation of the reactor operation, but was nevertheless proven to underestimate the MTC by a factor of 2 to 5. In this thesis, it is shown, both theoretically and experimentally, that the reason of the MTC underestimation by noise analysis is the radially loosely coupled character of the moderator temperature noise throughout the core. A new MTC noise estimator, accounting for this radially non-homogeneous moderator temperature noise is proposed and demonstrated to give the correct MTC value. This new MTC noise estimator relies on the neutron noise measured in a single point of the reactor and the radially averaged moderator temperature noise measured inside the core. In the case of the Ringhals-2 PWR in Sweden, Gamma-Thermometers (GTs) offer such a possibility since in dynamic mode they measure the moderator temperature noise, whereas in static

  19. Plutonium multi-recycling in increased moderating ratio reactors (IMR)

    International Nuclear Information System (INIS)

    Barbrault, P.; Larderet, P.

    1998-01-01

    The large core of the future jointly defined European PWR (EPR), would be compatible with an increased Moderating Ratio (MR) enabling better plutonium burnout. The purpose of current work on the subject is to assess plutonium multi-recycling possibilities in IMR reactors. What additional operating constraints would be involved under normal and accidental conditions and are they acceptable? The conclusion is that Plutonium multi-recycling in a PWR of the type envisaged for the EPR raises no major problems under the following conditions: use of an IMR MOX core, enhancing both plutonium burnout and absorber efficiency; use of enriched boron in both the primary coolant soluble boron and the B4C boron carbide in the control rods. Deeper investigation should be performed concerning the partial or total core drain-out, in view of the high total Pu concentrations involved (13%) and the types of core considered (100% MOX). (author)

  20. In-core LOCA (PTR) analysis with poisoned moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

    2005-01-01

    CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

  1. Liquid wall boiler and moderator (BAM) for heavy ion-pellet fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.R.; Lazareth, O.; Fillo, J.

    1977-11-01

    Thick liquid wall blankets appear to be of great promise for heavy ion pellet fusion reactors. They avoid the severe problems of intense radiation and blast damage that would be encountered with solid blanket structures. The liquid wall material can be chosen so that its vapor pressure at the working temperature of the power cycle is well below the value at which it might interfere with the propagation of the heavy ion beam. The liquid wall can be arranged so that it does not contact any surrounding solid structure when the pellet explosion occurs, including the ends. The ends can be magnetically closed just before the pellet explosion, or a time phased flow can be used, which will leave a clear central zone into which the pellet is injected. Parametric analysis comparing three candidate liquid wall materials were carried out. The three materials were lithium, flibe, and lead (with a low concentration of disolved lithium). Lead appeared to be the best choice for the liquid wall, although any of the three should allow a practical reactor system. The parametric analyses examined the effects of pellet yield (0 to 10 GJ), pellet mass (3 g to 3 kg), liquid wall thickness (10 cm to 80 cm), vapor condensation time (0 to 10 milliseconds), degree of neutron moderation in the pellet (none to 100%), liquid wall chamber size (radius of 1.5 meters to 4 meters), Pb/Li 6 ratio (100 to 5,000), and thickness of graphite moderating zone behind the liquid wall

  2. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

    1991-12-01

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  3. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  4. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  5. Inherent reactor power controller for a metal-fueled ALMR

    International Nuclear Information System (INIS)

    Wood, R.T.; Wilson, T.L. Jr.

    1990-01-01

    Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

  6. Automated scoping methodology for liquid metal natural circulation small reactor

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2014-01-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  7. An overview of IPPE research on liquid metal fast reactor thermohydraulics

    International Nuclear Information System (INIS)

    Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.

    2003-01-01

    The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model

  8. Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

    International Nuclear Information System (INIS)

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

  9. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Wade, D.C.

    1990-01-01

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  10. Graphite moderated {sup 252}Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Sajo B, L.; Barros, H.; Greaves, E. D. [Universidad Simon Bolivar, Nuclear Physics Laboratory, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a {sup 252}Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the {sup 252}Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  11. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    2007-12-01

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  12. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  13. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1990-01-01

    The United States have made substantial progress in achieving Advanced Liquid Metal Reactor (ALMR) program objectives. A decision was made in 1988 to select the General Electric ALMR concept known as PRISM (Power Reactor Innovative Safe Module) for advanced conceptual design. A 3-year contract was awarded to General Electric in January of last year for concentrated trade-off studies and advanced design development. The strategy is to integrate those advancements that best meet program objectives into a national ALMR system concept. (author). 10 figs, 1 tab

  14. Method of shielding a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    Sayre, R.K.

    1978-01-01

    The primary heat transport system of a nuclear reactor - particularly for a liquid-metal-cooled fast-breeder reactor - is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of a the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system

  15. Inherent safe fast breeder reactors and actinide burners, metallic fuel

    International Nuclear Information System (INIS)

    Dorner, S.; Schumacher, G.

    1991-04-01

    Nuclear power without breeder strategy uses the possibilities for the energy supply only to a small extend compared to the possibilities of fast breeder reactors, which offer an energy supply for thousands of years. Moreover, a fast neutron device offers the opportunity to run an actinide-burner that could improve the situation of waste management. Within this concept metallic fuel could play a key role. The present report shows some important aspects of the concept like the pyrometallic reprocessing, the behaviour of metallic fuel during a core meltdown accident and others. The report should contribute to the discussion of these problems and initialize further work

  16. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    International Nuclear Information System (INIS)

    P. Bernot

    2001-01-01

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% 235 U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to

  17. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2001-02-27

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited

  18. Moderator Chemistry Program

    International Nuclear Information System (INIS)

    Dewitt, L.V.; Gibbs, A.; Lambert, D.P.; Bohrer, S.R.; Fanning, R.L.; Houston, M.W.; Stinson, S.L.; Deible, R.W.; Abdel-Khalik, S.I.

    1990-11-01

    Over the past fifteen months, the Systems Chemistry Group of the Reactor Engineering Department has undertaken a comprehensive study of the Department's moderator chemistry program at Savannah River Site (SRS). An internal review was developed to formalize and document this program. Objectives were as outlined in a mission statement and action plan. In addition to the mission statement and action plan, nine separate task reports have been issued during the course of this study. Each of these task reports is included in this document as a chapter. This document is an organized compilation of the individual reports issued by the Systems Chemistry Group in assessment of SRS moderator chemistry to determine if there were significant gaps in the program as ft existed in October, 1989. While these reviews found no significant gaps in that mode of operation, or any items that adversely affected safety, items were identified that could be improved. Many of the items have already been dear with or are in the process of completion under this Moderator Chemistry Program and other Reactor Restart programs. A complete list of the items of improvement found under this assessment is found in Chapter 9, along with a proposed time table for correcting remaining items that can be improved for the chemistry program of SRS reactors. An additional external review of the moderator chemistry processes, recommendations, and responses to/from the Reactor Corrosion Mitigation Committee is included as Appendix to this compilation

  19. Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

    2006-02-01

    The primary components such as a reactor core, heat exchangers, pumps and internal structures of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A 0 plate wave was selected as the application mode of the sensor. The A 0 plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A 0 mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote inspection

  20. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  1. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  2. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  3. Dispersion strengthened ferritic alloy for use in liquid-metal fast breeder reactors (LMFBRS)

    International Nuclear Information System (INIS)

    Fischer, J.J.

    1978-01-01

    A dispersion-strengthened ferritic alloy is provided which has high-temperature strength and is readily fabricable at ambient temperatures, and which is useful as structural elements of liquid-metal fast breeder reactors. 4 tables

  4. Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

    International Nuclear Information System (INIS)

    Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

    2015-01-01

    At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

  5. Seismic analysis of a large pool-type LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Wang, C.Y.; Gvildys, J.

    1989-01-01

    This paper describes the seismic study of a 450-MWe liquid metal reactor (LMR) under 0.3-g SSE ground excitation. Two calculations were performed using the new design configuration. They deal with the seismic response of the reactor vessel, the guard vessel and support skirt, respectively. In both calculations, the stress and displacement fields at important locations of those components are investigated. Assessments are also made on the elastic and inelastic structural capabilities for other beyond-design basis seismic loads. Results of the reactor vessel analysis reveal that the maximum equivalent stress is only about half of the material yield stress. For the guard vessel and support skirt, the stress level is very small. Regarding the analysis if inelastic structural capability, solutions of the Newmark-Hall ductility modification method show that the reactor vessel can withstand seismics with ground ZPAs ranging from 1.015 to 1.31 g, which corresponds to 3.37 to 4.37 times the basic 0.3-g SSE. Thus, the reactor vessel and guard vessel are strong enough to resist seismic loads. 4 refs., 10 figs., 5 tabs

  6. Beneficial Re-Use of Metal from Decommissioning of Power Reactors

    International Nuclear Information System (INIS)

    Eshleman, Troy; Raw, Graham; Moloney, Barry

    2014-01-01

    Utilities and contractors decommissioning nuclear power reactors can recycle a high proportion of the scrap metal generated during dismantling either by free release for general re-use directly from the point of generation, or by recycling off-site at facilities specifically licensed for radioactive material. The worldwide commercial vendors operate different commercial models of volumetric decontamination of ferrous metals by thermal treatment. Some aim to achieve free release of output metals for general use, while others accept higher activity metals as feedstock for the manufacture of steel products which contain residual radioactivity, which we term 'Beneficial Re-use'. It is estimated that 10-30% of metals from light water reactor decommissioning have been exposed to neutron radiation (activated) and/or are contaminated to such an extent that free release is not achievable. This paper outlines a cost-effective alternative to managed storage or disposal for lightly activated or contaminated metal, utilising a 'Beneficial Re-Use' programme which has been in routine operation in the United States for over 20 years. 'Beneficial Re-Use' describes the manufacture of products such as radiation shielding from radioactive scrap metal. Unlike recycling practised in Europe, such products remain under control in licensed facilities and the metal does not find its way into general circulation or consumer products. Since 1992, EnergySolutions and its predecessor Duratek has been melting scrap at their Bear Creek, Tennessee facility to produce shield blocks for use in high energy research facilities. Over 62,300 t of scrap steel have been re-used, and the demand for shielding products continues long into the future. 3,000 t of this feedstock originated outside the US. This paper proposes the potential for activated steel that will not be acceptable at European recycling facilities to enter the Beneficial Re-use programme. Acceptance criteria

  7. Bioleaching of metals from spent refinery petroleum catalyst using moderately thermophilic bacteria: effect of particle size.

    Science.gov (United States)

    Srichandan, Haragobinda; Singh, Sradhanjali; Pathak, Ashish; Kim, Dong-Jin; Lee, Seoung-Won; Heyes, Graeme

    2014-01-01

    The present work investigated the leaching potential of moderately thermophilic bacteria in the recovery of metals from spent petroleum catalyst of varying particle sizes. The batch bioleaching experiments were conducted by employing a mixed consortium of moderate thermophilic bacteria at 45°C and by using five different particle sizes (from 45 to >2000 μm) of acetone-washed spent catalyst. The elemental mapping by FESEM confirmed the presence of Al, Ni, V and Mo along with sulfur in the spent catalyst. During bioleaching, Ni (92-97%) and V (81-91%) were leached in higher concentrations, whereas leaching yields of Al (23-38%) were found to be lowest in all particle sizes investigated. Decreasing the particle size from >2000 μm to 45-106 μm caused an increase in leaching yields of metals during initial hours. However, the final metals leaching yields were almost independent of particle sizes of catalyst. Leaching kinetics was observed to follow the diffusion-controlled model showing the linearity more close than the chemical control. The results of the present study suggested that bioleaching using moderate thermophilic bacteria was highly effective in removing the metals from spent catalyst. Moreover, bioleaching can be conducted using spent catalyst of higher particle size (>2000 μm), thus saving the grinding cost and making process attractive for larger scale application.

  8. Investigation of the noble metal deposition behaviour in boiling water reactors - the NORA project

    International Nuclear Information System (INIS)

    Ritter, S.; Karastoyanov, V.; Abolhassani-Dadras, S.; Guenther-Leopold, I.; Kivel, N.

    2010-01-01

    NobleChem™ is a technology developed by General Electric to reduce stress corrosion cracking (SCC) in reactor internals and recirculation pipes of boiling water reactors (BWRs) while preventing the negative side effects of classical hydrogen water chemistry. Noble metals (Pt, Rh) acting as electrocatalysts for the recombination of O 2 and H 2 O 2 with H 2 to H 2 O and thus reducing the corrosion potential more efficiently are injected into the feed water during reactor shut-down (classical method) or on-line during power operation. They are claimed to deposit as very fine metallic particles on all water-wetted surfaces including the most critical regions inside existing cracks and to stay electrocatalytic over long periods of time. The effectiveness of this technology in plants remains still to be demonstrated. Based on highly credible laboratory experiments down to the sub-ppb Pt concentration range, SCC mitigation may be expected, provided that a stoichiometric excess of H 2 and a sufficient surface coverage with very fine Pt particles exist simultaneously at the critical locations. Very little is known about the deposition and (re-)distribution behaviour of the Pt in the reactor. For the validation of this technique the research project NORA (noble metal deposition behaviour in BWRs) has been started at PSI with two main objectives: (i) to gain phenomenological insights and a better basic understanding of the Pt distribution and deposition behaviour in BWRs; (ii) to develop and qualify a non-destructive technique to characterise the size and distribution of the Pt particles and its local concentration on reactor components. This paper presents the objectives of the project, the planned work and a brief description of the status of the project. (author)

  9. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  10. Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

    1991-07-01

    A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

  11. Nuclear reactor pressure vessel with an inner metal coating covered with a high temperature resistant thermal insulator

    International Nuclear Information System (INIS)

    1974-01-01

    The thermal insulator covering the metal coating of a reactor vessel is designed for resisting high temperatures. It comprises one or several porous layers of ceramic fibers or of stacked metal foils, covered with a layer of bricks or ceramic tiles. The latter are fixed in position by fasteners comprising pins fixed to the coating and passing through said porous layers and fasteners (nut or bolts) for individually fixing the bricks to said pins, whereas ceramic plugs mounted on said bricks or tiles provide for the thermal insulation of the pins and of the nuts or bolts; such a thermal insulation can be applied to high-temperature reactors or to fast reactors [fr

  12. Liquid metal reactor development. Development of LMR coolant technology

    Energy Technology Data Exchange (ETDEWEB)

    Nam, H. Y.; Choi, S. K.; Hwang, J. s.; Lee, Y. B.; Choi, B. H.; Kim, J. M.; Kim, Y. G.; Kim, M. J.; Lee, S. D.; Kang, Y. H.; Maeng, Y. Y.; Kim, T. R.; Park, J. H.; Park, S. J.; Cha, J. H.; Kim, D. H.; Oh, S. K.; Park, C. G.; Hong, S. H.; Lee, K. H.; Chun, M. H.; Moon, H. T.; Chang, S. H.; Lee, D. N.

    1997-07-15

    Following studies have been performed during last three years as the 1.2 phase study of the mid and long term nuclear technology development plan. First, the small scale experiments using the sodium have been performed such as the basic turbulent mixing experiment which is related to the design of a compact reactor, the flow reversal characteristics experiment by natural circulation which is necessary for the analysis of local flow reversal when the electromagnetic pump is installed, the feasibility test of the decay heat removal by wall cooling and the operation of electromagnetic pump. Second, the technology of operation mechanism of sodium facility is developed and the technical analysis and fundamental experiments of sodium measuring technology has been performed such as differential pressure measuring experiment, local flow rate measuring experimenter, sodium void fraction measuring experiment, under sodium facility, the free surface movement experiment and the side orifice pressure drop experiment. A new bounded convection scheme was introduced to the ELBO3D thermo-hydraulic computer code designed for analysis of experimental result. A three dimensional computer code was developed for the analysis of free surface movement and the analysis model of transmission of sodium void fraction was developed. Fourth, the small scale key components are developed. The submersible-in-pool type electromagnetic pump which can be used as primary pump in the liquid metal reactor is developed. The SASS which uses the Curie-point electromagnet and the mock-up of Pantograph type IVTM were manufactured and their feasibility was evaluated. Fifth, the high temperature characteristics experiment of stainless steel which is used as a major material for liquid metal reactor and the material characteristics experiment of magnet coil were performed. (author). 126 refs., 98 tabs., 296 figs.

  13. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  14. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  15. Metal fire implications for advanced reactors. Part 1, literature review

    International Nuclear Information System (INIS)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-01-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior

  16. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor.

    Directory of Open Access Journals (Sweden)

    Shohreh Azizi

    Full Text Available For the effective application of a modified packed bed biofilm reactor (PBBR in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l concentrations of combined heavy metals at an optimum HRT condition (2 hours, while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l, is capable of removing the industrial contamination in wastewater.

  17. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor

    Science.gov (United States)

    Azizi, Shohreh; Kamika, Ilunga; Tekere, Memory

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l) concentrations of combined heavy metals at an optimum HRT condition (2 hours), while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD) of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l), is capable of removing the industrial contamination in wastewater. PMID:27186636

  18. Thermo-hydraulic test of the moderator cell of liquid hydrogen cold neutron source for the Budapest research reactor

    International Nuclear Information System (INIS)

    Grosz, Tamas; Rosta, Laszlo; Hargitai, Tibor; Mityukhlyaev, V.A.; Serebrov, A.P.; Zaharov, A.A.

    1999-01-01

    Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)

  19. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  20. Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

    1994-01-01

    Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

  1. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  2. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

  3. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

  4. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

  5. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  6. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  7. Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

    1991-01-01

    A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

  8. Failed fuel identification techniques for liquid-metal cooled reactors

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

    1995-01-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

  9. Toward a sustainable energy supply with reduced environmental burden. Development of metal fuel fast reactor cycle

    International Nuclear Information System (INIS)

    Koyama, Tadafumi; Kobayashi, Hiroaki; Kinoshita, Kensuke

    2009-01-01

    CRIEPI has been studying the metal fuel fast reactor cycle as an outstanding alternative for the future energy sources. In this paper, development of the metal fuel cycle is reviewed in the view point of technological feasibility and material balance. Preliminary estimation of reduction of the waste burden due to introduction of the metal fuel cycle technology is also reported. (author)

  10. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  12. Study on neutronics performance of flower shape advanced supercritical water cooled fast reactor with different solid moderators

    International Nuclear Information System (INIS)

    Yu Tao; Li Zhifeng; Xie Jinsen; Peng Honghua

    2015-01-01

    The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. K_e_f_f, conversion ratio, moderator temperature effect, Doppler effect and void effect of different material such as ZrH_1_._7, Bp, BeO, C and SiC are discussed. BeO and SiC hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel. (authors)

  13. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  14. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  15. RA Research reactor, Part I: Technical and operational properties of the RA reactor; Analiza sigurnosti rada Reaktora RA I-III, Deo I: Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Zecevic, V; Nikolic, M; Popovic, B; Milosevic, M; Milic, M; Strugar, P; Pesic, M; Nikolic, V; Rajic, M; Radivojevic, J; Jankovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    RA reactor is a research reactor with rather high power density. Apart from research it is used for isotope production and industrial applications due to high reactivity excess (about 11%). It is a thermal reactor, heavy water moderated, cooled by D{sub 2}O, and H{sub 2}O, with a graphite reflector. Nominal power is 6.5 MW. Fuel is 2% enriched metal uranium, reactor core height is 1220 mm, and diameter is 1405 mm. Reactor lattice is square with lattice pitch 130 mm. There is 6 horizontal experimental channels and a graphite column. There is a total of 84 fuel channels and 45 experimental channels in the core. Maximum thermal neutron flux is 5.5 10{sup 13} n/cm{sup 2} s at nominal power level.

  16. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  17. Hydrogen permeation resistant layers for liquid metal reactors

    International Nuclear Information System (INIS)

    McGuire, J.C.

    1980-03-01

    Reviewing the literature in the tritium diffusion field one can readily see a wide divergence in results for both the response of permeation rate to pressure, and the effect of oxide layers on total permeation rates. The basic mechanism of protective oxide layers is discussed. Two coatings which are less hydrogen permeable than the best naturally occurring oxide are described. The work described is part of an HEDL-ANL cooperative research program on Tritium Permeation in Liquid Metal Cooled Reactors. This includes permeation work on hydrogen, deuterium, and tritium with the hydrogen-deuterium research leading to the developments presented

  18. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  19. Primary Damage Characteristics in Metals Under Irradiation in the Cores of Thermal and Fast Reactors

    International Nuclear Information System (INIS)

    Pechenkin, V.A.

    2012-01-01

    For an analysis and forecasting of radiation-induced phenomena in structural materials of WWERs, PWRs and BN reactors the fast neutron fluence is usually used (for structural materials of the reactor cores and internals the fluence of neutrons with energy > 0.1 MeV, for WWER and PWRs vessel steels the fluence of neutrons with energy > 0.5 MeV in Russia and East Europe, and with energy > 1.0 MeV in USA and France). Displacements per atom (dpa) seem to be a more appropriate correlation parameter, because it allows comparing the results of materials irradiation in different neutron energy spectra or with different types of particles (neutrons, ions, fast electrons). Energy spectra of primary knocked atoms (PKA) and 'effective' dpa, which are introduced to take into account the point defect recombination during the relaxation stage of a displacement cascade, can be still better representation of the effect of irradiation on material properties. In this work the results of calculating dose rates (dpa/s, NRT-model), PKA energy spectra and PKA mean energies in metals under irradiation in the cores of Russian reactors WWER-440, WWER-1000 (both power thermal reactors) and BN-600 (power fast reactor) and BR-10 (test fast reactor) are presented. In all the reactors Fe and Zr are considered, with addition of Ti and W in BN-600. 'Effective' dose rates in these metals are calculated. Limitations and uncertainties in the standard dpa formulation (the NRT-dpa) are discussed. IPPE activities in the fields related to the TM subject are considered

  20. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    Science.gov (United States)

    Sibille, Laurent; Dominguez, Jesus A.

    2012-01-01

    The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

  1. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  2. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    International Nuclear Information System (INIS)

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop

  3. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    International Nuclear Information System (INIS)

    Kansal, Anuj Kumar; Joshi, Jyeshtharaj B.; Maheshwari, Naresh Kumar; Vijayan, Pallippattu Krishnan

    2015-01-01

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated

  4. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2015-06-15

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  5. Testing of seismic isolation bearings for advanced liquid metal reactor prism

    International Nuclear Information System (INIS)

    Tajirian, F.F.; Kelly, J.M.

    1988-01-01

    Seismic isolation can significantly mitigate earthquake loads on liquid metal reactors (LMR), thus reducing the impact of seismic loads on design. This improves plant safety margins for beyond-design basis seismic events and enhances adaptability of a standardized design to a variety of sites, with potential cost benefits. The PRISM (Power Reactor Inherently Safe Module) LMR incorporates a horizontal isolation system which consists of high damping steel laminated rubber bearings. The results of an experimental program to determine the mechanical properties of the rubber compound and the bearing performance under different loading conditions are presented. The test results demonstrate the excellent performance of the bearings and their suitability for isolating compact LMR plants

  6. Status of development and licensing support for advanced liquid metal reactors in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, D R [Argonne National Laboratory, Argonne, IL (United States); Gyorey, G [General Electric, San Jose, CA (United States)

    1991-07-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  7. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment

  8. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  9. Thermal Energetic Reactor with High Reproduction of Fission Materials

    International Nuclear Information System (INIS)

    Kotov, V.M.

    2012-01-01

    Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  10. Electrochemical membrane reactor: In situ separation and recovery of chromic acid and metal ions

    International Nuclear Information System (INIS)

    Khan, Jeeshan; Tripathi, Bijay P.; Saxena, Arunima; Shahi, Vinod K.

    2007-01-01

    An electrochemical membrane reactor with three compartments (anolyte, catholyte and central compartment) based on in-house-prepared cation- and anion-exchange membrane was developed to achieve in situ separation and recovery of chromic acid and metal ions. The physicochemical and electrochemical properties of the ion-exchange membrane under standard operating conditions reveal its suitability for the proposed reactor. Experiments using synthetic solutions of chromate and dichromate of different concentrations were carried out to study the feasibility of the process. Electrochemical reactions occurring at the cathode and anode under operating conditions are proposed. It was observed that metal ion migrated through the cation-exchange membrane from central compartment to catholyte and OH - formation at the cathode leads to the formation of metal hydroxide. Simultaneously, chromate ion migrated through the anion-exchange membrane from central compartment to the anolyte and formed chromic acid by combining H + produced their by oxidative water splitting. Thus a continuous decay in the concentration of chromate and metal ion was observed in the central compartment, which was recovered separately in the anolyte and catholyte, respectively, from their mixed solution. This process was completely optimized in terms of operating conditions such as initial concentration of chromate and metal ions in the central compartment, the applied cell voltage, chromate and metal ion flux, recovery percentage, energy consumption, and current efficiency. It was concluded that chromic acid and metal ions can be recovered efficiently from their mixed solution leaving behind the uncharged organics and can be reused as their corresponding acid and base apart from the purifying water for further applications

  11. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  12. Novel Composite Hydrogen-Permeable Membranes for Nonthermal Plasma Reactors for the Decomposition of Hydrogen Sulfide

    Energy Technology Data Exchange (ETDEWEB)

    Morris Argyle; John Ackerman; Suresh Muknahallipatna; Jerry Hamann; Stanislaw Legowski; Gui-Bing Zhao; Sanil John; Ji-Jun Zhang; Linna Wang

    2007-09-30

    The goal of this experimental project was to design and fabricate a reactor and membrane test cell to dissociate hydrogen sulfide (H{sub 2}S) in a nonthermal plasma and to recover hydrogen (H{sub 2}) through a superpermeable multi-layer membrane. Superpermeability of hydrogen atoms (H) has been reported by some researchers using membranes made of Group V transition metals (niobium, tantalum, vanadium, and their alloys), but it was not achieved at the moderate pressure conditions used in this study. However, H{sub 2}S was successfully decomposed at energy efficiencies higher than any other reports for the high H{sub 2}S concentration and moderate pressures (corresponding to high reactor throughputs) used in this study.

  13. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    International Nuclear Information System (INIS)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

    1998-08-01

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

  14. Fine distributed moderating material to the enhance feedback effects in LBE cooled rast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    In this work it is demonstrated, that the concept of enhanced feedback coefficients is transferable to LBE cooled fast reactors. The demonstration is based on the fuel assembly design of the CDT project. The effect of the moderating material on the neutron spectrum, on the k{sub inf}, and on the fuel temperature feedback and the coolant feedback is shown, discussed and compared to SFRs. The calculations are performed with the 2D lattice transport code HELIOS and based on the fully detailed fuel assembly geometry representation. (orig.)

  15. Metal fires and their implications for advanced reactors. Part 3: Experimental and modeling results

    International Nuclear Information System (INIS)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-01-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  16. Potential applications of robotics in advanced liquid-metal reactors

    International Nuclear Information System (INIS)

    Carroll, D.G.; Thompson, M.L.

    1990-01-01

    The advanced liquid-metal reactor (ALMR) design includes a range of robots and automation devices. They extend from stationary robots that are a part of the current design to more exotic concepts with mobile, autonomous units, which may become part of the design. Development of robotic application requirements is enhanced by using computer models of work spaces in three dimensions. The primary goals of the more autonomous machines are to: (1) extent and/or enhance one's capabilities in a hazardous environment; some tasks could encounter high temperatures (up to 800 degree F), high radiation (fields up to several hundred thousand roentgens per hour), rooms filled with inert gas and/or sodium aerosol, or combinations of these; (2) reduce operating and maintenance cost through inservice inspection (ISI) of various parts of the reactor, through consideration of as-low-as-reasonably achievable radiation levels, and through automation of some maintenance/processing operations. This paper discusses some applications in the fuel cycle, in refueling operations, and in inspection

  17. Status of Liquid Metal Fast Reactor Development in the United States of America, March 1987

    International Nuclear Information System (INIS)

    Horton, K.E.

    1987-01-01

    In order to meet the objective to develop and demonstrate economically competitive reactor designs and associated fuel cycles early in the next century, the U.S. program has become more focused. Two innovative reactor designs supported by the metal-fueled Integral Fast Reactor program are being directed at fulfilling a series of advanced reactor goals. The supporting technology programs and facilities are being refocused to support the overall goals. International collaboration is being broadened to provide the two-way support across the spectrum of plant projects and the fuel cycle. This program is intended to maintain the technology base into the time period (mid-1990s) when a private sector demonstration could be initiated. (author)

  18. Transmutation of Tc-99 in fission reactors

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  19. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1996-04-01

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

  20. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs.

  1. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  2. Moderator mixing after a pressure tube failure

    International Nuclear Information System (INIS)

    MacKinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

    1997-01-01

    During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system conditions investigated are of a reactor in a GSS, with coolant in the primary heat transport system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The

  3. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  4. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1989-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactor and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long duration earthquakes. Seismic isolation is accomplished with high damping natural rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program is described in this paper and selected results are presented. The initial testing indicated excellent performance of high damping natural rubber bearings

  5. Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2001-01-01

    In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

  6. Development of a risk-based inservice inspection program for a liquid metal reactor

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.

    1996-01-01

    The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

  7. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Doshi, P.K.; George, R.A.; Dollard, W.J.

    1982-01-01

    A mechanical spectral shift arrangement for controlling a nuclear reactor includes a plurality of reactor coolant displacer members which are inserted into a reactor core at the beginning of the core life to reduce the volume of reactor coolant-moderator in the core at start-up. However, as the reactivity of the core declines with fuel depletion, selected displacer members are withdrawn from the core at selected time intervals to increase core moderation at a time when fuel reactivity is declining. (author)

  8. Reactivity and reaction rate ratio changes with moderator voidage in a light water high converter reactor lattice

    International Nuclear Information System (INIS)

    Chawla, R.; Gmuer, K.; Hager, H.; Seiler, R.

    1986-01-01

    Integral reaction rate ratios and other k ∞ related parameters have been measured in the first three cores of the experimental program on light water high converter reactor (LWHCR) test lattices in the PROTEUS reactor. The reference tight-pitch lattice consisted of two rod types, with an average fissile-plutonium enrichment of 6% and a fuel/moderator ratio of 2.0. The moderators were H 2 O, Dowtherm (simulating an H 2 O voidage of 42.5%), and air (100% void). Comparisons of the measured parameters have been made with calculational results based mainly on the use of two separate codes and their associated data libraries, namely, WIMS-D and EPRI-CPM. A reconstruction of individual components of the k-infinity void coefficient has been carried out on the basis of the measured changes with voidage of the various reaction rate ratios, as well as of k-infinity itself. The subsequent more detailed comparisons between experiment and calculation should provide a useful basis for resolving the conflicting calculational results that have been reported in the past for the void coefficient characteristics of LWHCRs. (author)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  10. Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

    Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

  11. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  12. Experience with valves for PHWR reactors

    International Nuclear Information System (INIS)

    Narayan, K.; Mhetre, S.G.

    1977-01-01

    Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

  13. The role of metal complexes in nuclear reactor decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Prince, A.A.M.; Raghavan, P.S.; Gopalan, R. [Madras Christian College, Tambaram, Chennai (India); Velmurugan, S.; Narasimhan, S.V. [Bhabha Atomic Research Center (BARC) (IN). Water and Steam Chemistry Lab. (WSCL)

    2006-07-15

    Chemical decontamination is the process of removal of radioactivity from corrosion products formed on structural materials in the nuclear reactors. These corrosion products cause problems for the operation and maintenance of the plants. Removal of the radioactive contaminants can be achieved by dissolving the oxide from the system surface using organic complexing agents in low concentrations known as dilute chemical decontamination (DCD) formulations. These organic complexing agents attack the oxide surface and form metal complexes, which further accelerate the dissolution process. The stability of the complexes plays an important role in dissolving the radioactive contaminated oxides. In addition, the DCD process is operated through ion exchange resins for the removal of the dissolved metal ions and radioactive nuclides. In the present study, the kinetics of dissolution of various model corrosion products such as magnetite (Fe{sub 3}O{sub 4}), hematite ({alpha}-Fe{sub 2}O{sub 3}) and maghemite ({gamma}-Fe{sub 2}O{sub 3}) have been studied in the presence of complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), hydroxyethylethylenediaminepentaacetic acid (HEEDTA), and 2,6 pyridinedicarboxylic acid (PDCA). The reductive roles of metal complexes and organic reducing agents are discussed. (orig.)

  14. SWEEP, a computer program for the analysis of CDA energetics in liquid metal reactors

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

    2003-12-01

    The SWEEP computer code was developed in this study to evaluate the work energy arising from two-phase expansion of fuel or sodium during core disruptive accidents in KALIMER. In the SWEEP program, scoping calculations with a modified Bethe-Tait method is first carried out using SCHAMBETA module to provide the initial thermodynamic conditions for the subsequent analyses to estimate the mechanical work energy generated in the reactor system. To estimate the work energy due to fuel-vapor expansion, a bounding approach is adopted to calculate the work potential assuming isentropic expansion to atmospheric pressure during super-prompt critical power excursions. Work potentials are also calculated in the SWEEP code for sodium expansion using the simple thermodynamic models including the infinite heat transfer model during expansion(Hicks and Menzies method) or more realistic zero heat transfer model for a typical initial condition of core disruptive accident. Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool-type sodium cooled prototype fast reactor that uses U-TRU-Zr metallic fuel

  15. The effectivty of hydrogeneous moderators in pulsed sources

    International Nuclear Information System (INIS)

    Rief, H.; Hartman, J.

    1975-01-01

    Guide lines are provided for an evaluation of the potential of pulsed reactors. In the SORA reactor, neutrons emitted from the fast core are converted in hydrogeneous moderators to beams of low energy neutrons for time of flight experiments. The important characteristics of the neutron sources are absolute intensity of the neutron beam and its energy and time distribution. The problem is solved mathematcially by the random walk (Monte Carlo) method. Calculational methods which are described are compared with pulsed moderator measurements. The choice of moderators and criteria of optimization are discussed. Particular examples of realistic moderator design as planned for SOYA, and as they will be used in pulsed reactors, are analysed, a distinction being made between thermal, cold, and hot moderators. Finally flux estimates are compared with those obtained for a spallation target. (U.K.)

  16. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    Chang, Se Myong; Park, A. Y.; Kim, Hyoung Tae

    2011-01-01

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  17. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  18. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  19. All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

    International Nuclear Information System (INIS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

    2009-01-01

    Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

  20. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  1. Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Choi, Seok Ki; Kim, Seong O

    2011-01-01

    A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

  2. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  3. Physics considerations in the design of liquid metal reactors for transuranium element consumption

    International Nuclear Information System (INIS)

    Khalil, H.; Hill, R.; Fujita, E.; Wade, D.

    1992-01-01

    The management of transuranic nuclides in liquid metal reactors (LMR's) is considered based on the use of the Integral Fast Reactor (IFR) concept. Unique features of the IFR fuel cycle with respect to transuranic management are identified. These features are exploited together with the hard spectrum of LMR's to demonstrate the neutronic feasibility of a wide range of transuranic management options ranging from efficient breeding to pure consumption. Core physics aspects of the development of a low sodium void worth transuranic burner concept are described. Neutronics performance parameters and reactivity feedback characteristics estimated for this core concept are presented

  4. Metallic sodium as a coolant of high speed nuclear reactors, (2)

    International Nuclear Information System (INIS)

    Atsumo, Hideo

    1975-01-01

    Tables are given on all the sodium loops in Japan and most of the sodium loops all over the world. Name and purpose of the loops, time of establishment, highest temperature, amount of sodium, flow rate, the materials used for the construction of the loops, and the diameter of the main pipings are given. Also, the problems related with these loops are discussed. For example, the high temperature sodium facility at HEDL-WADCO was made for the FFTF component test and instrument test, and uses 50,000 gallons of metallic sodium. The highest temperature is 590 0 C. The sodium flows at the rate of 60 g/m. The body is made of Type 304 stainless steel. Main data of existing sodium-cooled reactors in the world are also tabulated. The data include thermal output, electric output, the structure of the reactor cores, the dimensions of the cores, fuel used, the highest temperature in the reactors, the temperature of sodium at the inlet and outlet, the rate of multiplication, the amount of sodium used, number of control rods, number of heat exchangers, and the pressure of steam. The Monju type nuclear reactor in Japan uses 1,800 ton of sodium. (Fukutomi, T.)

  5. Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

    International Nuclear Information System (INIS)

    Wootan, D.W.; Omberg, R.P.

    2015-01-01

    Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

  6. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

  7. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  8. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

  9. Radiation Damage in Reactor Materials. Part of the Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-08-15

    Radiation damage has presented a new design parameter for the selection of materials to be used in fuel and cladding elements, moderators, structural components and pressure vessels in nuclear reactors. The severe and novel requirements for certain optimum combinations of physical and nuclear properties have emphasized the need for a better understanding of the basic mechanisms of radiation damage. This knowledge is not only essential for progress in the field of nuclear energy, but has direct applications to space technology and semi-conductor research as well. The IAEA, as part of its programme of promoting nuclear technology, therefore convened the Symposium on Radiation Damage in Solids and Reactor Materials, 7-11 May 1962. At the invitation of, and with generous material assistance from, the Government of Italy, the Symposium was held at Venice. The Symposium was primarily concerned with the investigation of the fundamental processes of radiation that underlie the behaviour of metals, alloys and ceramics that are actually useful or potentially useful reactor materials. Two sessions were devoted to studies of irradiation effects on simple metals, as these effects are easiest to interpret. Other topics included general theory, alloys, fissionable and moderator materials and special experimental techniques for radiation damage studies. The properties influenced by irradiation which were of main concern were those of primary importance to the behaviour of solids as reactor materials (e. g. dimensional stability, phase transformation, radiation hardening, fracture, fission-gas escape from uranium and its compounds). Other properties, such as optical, electrical and magnetic properties, and effects on semiconductors, ionic and other non-metallic crystals are also of interest in that these studies can increase our knowledge of the mechanism of radiation damage in solids and provide a tool for investigation into the physics of the solid state by offering a means of

  10. Radiation Damage in Reactor Materials. Part of the Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials

    International Nuclear Information System (INIS)

    1963-01-01

    Radiation damage has presented a new design parameter for the selection of materials to be used in fuel and cladding elements, moderators, structural components and pressure vessels in nuclear reactors. The severe and novel requirements for certain optimum combinations of physical and nuclear properties have emphasized the need for a better understanding of the basic mechanisms of radiation damage. This knowledge is not only essential for progress in the field of nuclear energy, but has direct applications to space technology and semi-conductor research as well. The IAEA, as part of its programme of promoting nuclear technology, therefore convened the Symposium on Radiation Damage in Solids and Reactor Materials, 7-11 May 1962. At the invitation of, and with generous material assistance from, the Government of Italy, the Symposium was held at Venice. The Symposium was primarily concerned with the investigation of the fundamental processes of radiation that underlie the behaviour of metals, alloys and ceramics that are actually useful or potentially useful reactor materials. Two sessions were devoted to studies of irradiation effects on simple metals, as these effects are easiest to interpret. Other topics included general theory, alloys, fissionable and moderator materials and special experimental techniques for radiation damage studies. The properties influenced by irradiation which were of main concern were those of primary importance to the behaviour of solids as reactor materials (e. g. dimensional stability, phase transformation, radiation hardening, fracture, fission-gas escape from uranium and its compounds). Other properties, such as optical, electrical and magnetic properties, and effects on semiconductors, ionic and other non-metallic crystals are also of interest in that these studies can increase our knowledge of the mechanism of radiation damage in solids and provide a tool for investigation into the physics of the solid state by offering a means of

  11. Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    1978-06-01

    The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

  12. Physics of Plutonium Recycling in Thermal Reactors

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1967-01-01

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  13. Physics of Plutonium Recycling in Thermal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1967-09-15

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  14. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  15. A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

    International Nuclear Information System (INIS)

    Choi, H.; Downar, T.J.

    1999-01-01

    A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

  16. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    1989-01-01

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

  17. Pumps of molten metal based on magnetohydrodynamicprinciple for cooling high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Donátová, M.; Karban, P.; Ulrych, B.

    2009-01-01

    Roč. 85, č. 4 (2009), s. 13-15 ISSN 0033-2097 R&D Projects: GA ČR(CZ) GA102/07/0496 Institutional research plan: CEZ:AV0Z20570509 Keywords : pumps of molten metal * magnetohydrodynamic principle * nuclear reactors Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.196, year: 2009

  18. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1991-01-01

    An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Subsequent to the directive to shut down the Fast Flux Test Facility in early 1990, a variety of activities were undertaken to provide support for continued operation. The United States has made substantial progress in achieving ALMR program objectives. The metal fuel cycle is designed to recycle and burn its own actiniums, and has the potential to be a very effective burner of actiniums generated in the LWRs. The current emphasis in the IFR Program is on the comprehensive development of the IFR (Integral Fast Reactor) technology, to be followed by a period of technology demonstration which would verify the economic feasibility of the concept. The United States has been active in international cooperative activities in the fast reactor sector since 1969. (author). 11 figs, 1 tab

  19. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  20. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-01-01

    This current report is a summary of information obtained in the 'Information Capture' task of the U.S. DOE-funded 'Under Sodium Viewing (USV) Project.' The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  1. Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2014-01-01

    Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities

  2. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

  3. Assessment of a Pressure Tube Rupture with a Poisoned Moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, S. C.; Kim, E. K.

    2005-01-01

    The postulated in-core LOCA has been analyzed and evaluated while the reactor is operating normally with a low moderator poison concentration for CANDU. However, when the reactor is operating with a relatively large amount of boron and/or gadolinium poison in the moderator, an assessment of the fuel integrity was required for the pressure tube rupture (PTR) accident. Poisoned moderator exists mainly during a startup after a prolonged shutdown lasting for more than one day. For the case of a reactor regulating system (RRS) working, the methodology of the PTR assessment with a poisoned moderator has been developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for the Wolsong Nuclear Power Plants recently. The developed methodology and results are presented

  4. Fusion reactor pumped laser

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1988-01-01

    A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

  5. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  6. Moderation of neutron energy

    International Nuclear Information System (INIS)

    Marlatt, G.R.

    1986-01-01

    This patent describes a nuclear reactor system having a nuclear reactor which has a core including fuel assemblies, means for transmitting through the core a coolant, the coolant having a predetermined neutron-energy moderating property, sealed tubes in the core, each tube containing a material having a different neutron-energy moderating property than the coolant, means, when actuated, to engage at least certain of the tubes, for opening certain of the tubes to permit the coolant to replace the material in the tubes thereby to change the energy spectrum of the neutrons in the reactor, hydraulic means, connected to the opening means, for actuating the opening means to engage certain of the tubes to open the tubes. A device, external to the reactor, connected to the hydraulic means controlls the actuation of the opening means, the opening means being so set with reference to the tubes that only certain of the tubes are opened at any time as the opening means is advanced towards the tubes by the hydraulic means

  7. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  8. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-01-01

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  9. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

    2013-10-15

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  10. Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance

  11. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  12. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors

    International Nuclear Information System (INIS)

    Martin Munoz, M.; Alvarez Gonzalez, F.

    1969-01-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs

  13. Metal halides vapor lasers with inner reactor and small active volume.

    Science.gov (United States)

    Shiyanov, D. V.; Sukhanov, V. B.; Evtushenko, G. S.

    2018-04-01

    Investigation of the energy characteristics of copper, manganese, lead halide vapor lasers with inner reactor and small active volume 90 cm3 was made. The optimal operating pulse repetition rates, temperatures, and buffer gas pressure for gas discharge tubes with internal and external electrodes are determined. Under identical pump conditions, such systems are not inferior in their characteristics to standard metal halide vapor lasers. It is shown that the use of a zeolite halogen generator provides lifetime laser operation.

  14. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  15. Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    Gillett, J.E.; Wineman, A.L.

    1983-01-01

    A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns each have a pin which rides in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring the locking plates into engagement with the housing in a raised or a lowered position of the support column such that the support column is then supported by the locking plate and not by the ball screw jacks. (author)

  16. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1991-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US DOE. The objective of this development is to provide, by the turn of the century, a reactor with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US DOE, ANL, Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), GE, and Bechtel National, Inc. (BNI), is described and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team. (orig./HP)

  17. Continuous Polyol Synthesis of Metal and Metal Oxide Nanoparticles Using a Segmented Flow Tubular Reactor (SFTR

    Directory of Open Access Journals (Sweden)

    Andrea Testino

    2015-06-01

    Full Text Available Over the last years a new type of tubular plug flow reactor, the segmented flow tubular reactor (SFTR, has proven its versatility and robustness through the water-based synthesis of precipitates as varied as CaCO3, BaTiO3, Mn(1−xNixC2O4·2H2O, YBa oxalates, copper oxalate, ZnS, ZnO, iron oxides, and TiO2 produced with a high powder quality (phase composition, particle size, and shape and high reproducibility. The SFTR has been developed to overcome the classical problems of powder production scale-up from batch processes, which are mainly linked with mass and heat transfer. Recently, the SFTR concept has been further developed and applied for the synthesis of metals, metal oxides, and salts in form of nano- or micro-particles in organic solvents. This has been done by increasing the working temperature and modifying the particle carrying solvent. In this paper we summarize the experimental results for four materials prepared according to the polyol synthesis route combined with the SFTR. CeO2, Ni, Ag, and Ca3(PO42 nanoparticles (NPs can be obtained with a production rate of about 1–10 g per h. The production was carried out for several hours with constant product quality. These findings further corroborate the reliability and versatility of the SFTR for high throughput powder production.

  18. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  19. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    Chang, S.M.; Kim, H.T.

    2013-01-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  20. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

    1995-01-01

    Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

  1. Analysis of power variation in a CANDU-6 with a loss of moderator

    International Nuclear Information System (INIS)

    Fan, Y.

    2008-01-01

    A loss of heavy water in a postulated small failure in the horizontal unpressurized calandria vessel of a CANDU-6 reactor will lead to a drop in the moderator level in the reactor core. The STEPBACK and SETBACK functions at the initial moment of the drop in moderator level ensure a reactor shutdown and a reduction in total reactor power during this 900 seconds postulated transient. If the STEPBACK and SETBACK functions are unavailable, the reactor's regulating system will try to compensate for the negative reactivity resulting from the loss of the moderator. This kind of compensation will lead to power distortions from top to bottom in the reactor core. .Comparisons of different moderator leakage rates were used in the analysis to determine the relationships between the power and the moderator leakage rates. Maximum bundle and channel powers obtained were insensitive to the moderator leakage rate. .In a complete analysis for a moderator leakage rate of 40 1/s, it was found that, without the STEPBACK and SETBACK functions, serious power distortions would occur during the 900 seconds transient. The maximization of bundle and channel power during this transient happened in the bottom part of the reactor , and the regulating system worsened this power distortion. .From the above analysis, it was concluded that the maximum bundle power attained during the loss of the moderator was 1.18% of its initial value. The risk of bundle dryout was, therefore, quite small. (author)

  2. Inducer pumps for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Jackson, E.D.

    2002-01-01

    Pumps proposed for liquid metal reactor plants typically use centrifugal impellers as the rotating element and are required to maintain a relatively low speed to keep the suction specific speed low enough to operate at the available net positive suction head (HPSH) and to avoid cavitation damage. These low speeds of operation require that the pump diameter increase and/or multiple stages be used to achieve the design head. This frequently results in a large, heavy, complex pump design. In addition, the low speed results in a larger drive motor size so that the resultant penalty to the plant designer is multiplied. The heavier pump can also result in further complications as, e.g., the difficulty in maintaining the first critical speed sufficiently above the pump operating range to provide margin for rotor dynamic stability. To overcome some of these disadvantages, it was proposed the use of inducer pumps for Liquid Metal Fast Breeder Reactor (LMFBR) plants. This paper discusses some of the advantages of the inducer pump and the development history of designing and testing these pumps both in water and sodium. The inducer pump is seen to be a sound concept with a strong technology base derived from the aerospace and ship propulsion industries. The superior suction performance capability of the inducer offers significant system design advantages, primarily a smaller, lighter weight, less complex pump design with resulting saving in cost. Extensive testing of these pumps has been conducted in both sodium and water to demonstrate the long-life capability with no cavitation damage occurring in those designs based on Rockwell's current design criteria. These tests have utilized multiple inspection and measurement approaches to accurately assess and identify any potential for cavitation damage, and these approaches have all concluded that no damage is occurring. Therefore, it is concluded that inducer pumps can be safely designed for long life operation in sodium with

  3. Liquid metal versus gas cooled reactor concepts for a turbo electric powered space vehicle

    International Nuclear Information System (INIS)

    Carre, F.; Proust, E.; Schwartz, J.P.

    1985-01-01

    Recent CNES/CEA prospective studies of an orbit transfer vehicule to be launched by ARIANE V, emphasize the advantage of the Brayton cycle over the thermionics and thermoelectricity, in minimizing the total mass of 100 to 300 kWsub(e) power systems under the constraint specific to ARIANE of a radiator area limited to 95 m 2 . The review of candidate reactor concepts for this application, finally recommends both liquid metal and gas cooled reactors, for their satisfactory adaptation to a reference Brayton cycle and for the available experience from the terrestrial operation of comparable systems

  4. Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    2004-01-01

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

  5. Experience of on-site disposal of production uranium-graphite nuclear reactor.

    Science.gov (United States)

    Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

    2018-04-01

    The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

  6. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  7. Neutron flux measuring system for nuclear reactor

    International Nuclear Information System (INIS)

    Aoki, Kazuo.

    1977-01-01

    Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

  8. Reactor shutdown back-up system

    International Nuclear Information System (INIS)

    Hirao, Seizo; Sakashita, Motoaki.

    1982-01-01

    Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

  9. Demonstration test on manufacturing 200 l drum inner shielding material for recycling of reactor operating metal scrap

    International Nuclear Information System (INIS)

    Umemura, A.; Kimura, K.; Ueno, H.

    1993-01-01

    Low-level reactor wastes should be safely recycled considering those resource values, the reduction of waste disposal volume and environmental effects. The reasonable recycling system of reactor operating metal scrap has been studied and it was concluded that the 200 liter drum inner shielding material is a very promising product for recycling within the nuclear industry. The drum inner shielding material does not require high quality and so it is expected to be easily manufactured by melting and casting from roughly sorted scrap metals. This means that the economical scrap metal recycling system can be achieved by introducing it. Furthermore its use will ensure safety because of being contained in a drum. In order to realize this recycling system with the drum inner shielding material, the demonstration test program is being conducted. The construction of the test facility, which consists of a melting and refining furnace, a casting apparatus, a machining apparatus etc., was finishing in September, 1992

  10. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  11. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  12. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    International Nuclear Information System (INIS)

    Rosenthal, Murray Wilford

    2009-01-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  13. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

    The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

  14. Example of End States of Decommissioning Phases from the Decommissioning of the Multipurpose Research Reactor MZFR, Karlsruhe, Germany

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-06-15

    The multipurpose reactor MZFR was a pressurized water reactor, cooled and moderated with heavy water. It was built from 1961 to 1966, and went critical for the first time on 29 September 1965. After 19 years of successful operation, the reactor was shut down on 3 May 1984. The reactor had a thermal output of 200 MW, and an electrical output of 50 MW. In addition to generating electricity, the MZFR had the following functions: - Testing fuel assemblies and various materials for reactor construction; - Gaining experience in the design, erection and operation of heavy water reactor systems; - Training scientific and technical reactor personnel; - Providing heat (first nuclear combined heat and power system (1979-1984)). In 1989, it was decided to dismantle the reactor completely, step by step. The decommissioning concept for the plant, down to a greenfield site, provides for eight distinct decommissioning steps (phases). A separate decommissioning licence was required for each step. The decommissioning work was carried out according to pre-approved work schedules. About 72 000 t of concrete and 7200 t of metal were to be removed. About 1000 t of concrete (500 t biological shield) and 1680 t of metal were to be classified as radioactive waste.

  15. Experimental and design experience with passive safety features of liquid metal reactors

    International Nuclear Information System (INIS)

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  16. Metal Hall sensors for the new generation fusion reactors of DEMO scale

    Science.gov (United States)

    Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

    2017-11-01

    For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

  17. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  18. Italian position paper on the safety analysis of liquid metal fast breeder reactors as related to sodium fires. The PEC reactor

    International Nuclear Information System (INIS)

    Gerosa, A.

    1983-01-01

    To obtain a deep understanding of physical phenomena and engineering problems connected to sodium fires, and to optimize the utilization of human and financial resources available, CNEN (now ENEA) has decided to join the French Commissariat a l'Energie Atomique (CEA) in the realization of a Franco-Italian experimental programme on sodium fires, named ESMERALDA. As for design preventions for PEC reactor (a fast flux, liquid metal cooled, fuel element testing reactor) fundamental choices were made taking into account all available knowledge, but with particular reference to the results of CEA's previous experiments on sodium fires. More detailed design analysis will be possible in the future, based on experimental results coming from the ESMERALDA programme

  19. Study of MHD problems in liquid metal blankets of fusion reactors

    International Nuclear Information System (INIS)

    Michael, I.

    1984-12-01

    This study describes in a concise form the state of knowledge regarding MHD problems to be expected in case of use of liquid metal in the blankets of fusion reactors with magnetic confinement. MHD pressure losses and MHD friction coefficients in the straight channel, in bent sections and in case of variation of the channel cross section play a major role because the high MHD flow resistances call for high pumping powers. Influencing the velocity profile transverse to the main flow direction of the liquid metal by application of an external, strong magnetic field bears consequences on the release and transport of corrosion products in the liquid metal circuit and on the heat transfer. Possibilities of reducing the MHD effects are discussed. However, it becomes obvious that an account of the lack of experimental results there are still major gaps in the knowledge of MHD effects occurring in strong magnetic fields. These gaps can be greatly reduced by implementation of an experimental program as proposed in this report. (orig.) [de

  20. The role of structural integrity in liquid metal fast breeder reactor safety

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1982-01-01

    Extensive studies have demonstrated the favourable safety characteristics of liquid metal fast breeder reactors, which are attributable to both their inherent features and the engineered safeguards which are included. This requires demonstration that there is no risk of sudden catastrophic failure of the core support system allowing the core to drop off the control rods to give a prompt critical reactivity excursion. An important part of our work in support of the safety case for the U.K. Commercial Demonstration Fast Reactor is to demonstrate that such a failure is virtually incredible. This covers design features, study of the fracture behaviour of stainless steel structures, and inspection and monitoring during fabrication and service. The paper gives a broad description of the relevant design features and supporting work programme

  1. A safety concern related to CANDU moderator subcooling and status of KAERI moderator circulation test (MCT) experiments

    International Nuclear Information System (INIS)

    Rhee, Bo W.; Kim, Hyoung T.; Kim, Tongbeum; Im, Sunghyuk

    2015-01-01

    The flow inside the moderator tank of a CANDU-6 reactor during full power steady state operation has been suspected to be operating in the buoyancy/inertial driven mixed convection regime as illustrated in the middle figure. At some regions of the moderator tank where the buoyancy driven upward flow and the inertial momentum driven downward flows interface counter-currently, there exist some interface regions between these two flows like the middle one, and the local temperatures at these interface regions are known to oscillate with different amplitude at various fluctuation frequencies as shown. According to a numerical simulation of the moderator flow and temperature distribution at full power steady state carried out by previous researches showed that any small disturbances in the flow or temperature may initiate the system unstable and aggravate the asymmetric flow and temperature patterns. The tests at the 3-D Moderator Test Facility (MTF) that is a representative scaled-down of CANDU reactors, reproduced the expected and observed moderator behavior in the reactor as well as the local temperature fluctuations arising from the delicate balance of forced and buoyancy induced flow. This observation raised a safety concern as the local moderator temperature at some regions showed fluctuations with an amplitude that may jeopardize the safety margin, i.e. the difference between the available subcooling and the subcooling requirement. The scope of this paper is to review the basis of the safety concern related to this moderator subcooling and local temperature fluctuation and describe the current status of MCT erection and some of the experiments carried so far

  2. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  3. Electrorefining open-quotes Nclose quotes reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.

    1995-01-01

    Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined

  4. Control rod for the operation of nuclear reactor

    International Nuclear Information System (INIS)

    Ishida, Hiromi

    1987-01-01

    Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

  5. Initial charge reactor core

    International Nuclear Information System (INIS)

    Kiyono, Takeshi

    1984-01-01

    Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

  6. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  7. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  8. An Overview of Ageing Management and Refurbishment of Research Reactors at Trombay

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, R. C.; Raina, V. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2014-08-15

    Three nuclear research reactors have been in operation at Bhabha Atomic Research Centre, Mumbai, India. India has a rich experience of about 120 research reactor operating years including ageing management. A well structured programme is in force for plant life management, refurbishment and upgrading reactors in operation. Apsara, commissioned in August 1956, was the first research reactor. Apsara is a 1 MW{sub th} swimming pool type of reactor with a movable core loaded with enriched uranium fuel and immersed in demineralized light water pool, which serves as coolant, moderator and reflector besides providing radiation shielding. Apsara was shut down during May 2009 for partial decommissioning and upgrading to a 2 MW reactor with several safety upgrades, e.g. a LEU based reactor core with higher neutron flux, a new reactor building meeting seismic qualification criteria and two independent shutdown devices. Cirus, a 40 MW{sub th} tank type reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; has been in operation since 1960. After about three decade of operation, the availability factor started declining mainly due to outage of equipment exhibiting signs of ageing. After ageing studies and performance review, refurbishment requirements were identified. A programme for refurbishment was drawn that included safety upgrades like civil repairs to the emergency storage reservoir to meet seismic qualification criteria and a new iodine removal system for better efficiency. The reactor was shut down during 1997 for execution of this refurbishment programme. After completion of refurbishment, the reactor was brought back into operation during 2003. It has completed about seven years of safe operation after refurbishment with a significant increase in availability factor from 70% to about 90%. The reactor was permanently shut down during December 2010. The reactor core was unloaded

  9. Plans for the development of the IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    Johnson, T.R.

    1986-01-01

    The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium

  10. A Polyethylene Moderator Design for Auxiliary Ex-core Neutron Detector

    International Nuclear Information System (INIS)

    Lee, Hwan Soo; Shin, Ho Cheol; Bae, Seong Man

    2012-01-01

    The moderator of detector assembly in ENFMS (Excore Neutron Flux Monitoring System) plays a key role for slowing down from fast neutron to thermal neutron at outside of reactor vessel. Since neutron monitoring detector such as BF3, fission chamber detectors mostly responds to thermal neutron, moderator should be included to neutron detector assembly to detect more efficiently. Generally, resin has been used for moderator of detector in ENFMS of OPR1000 and APR1400, because resin has stable thermal resistance, availability and high neutron moderation characteristics due to the light atomic materials. In case of an auxiliary ex-core neutron detector, the polyethylene is suggested that polyethylene has a better moderator rather than resin, then, the amounts of moderator are reduced. This is important thing for auxiliary ex-core detector equipment at reactor, because the auxiliary equipment should affect minimally to another system. In this study, polyethylene moderator is designed for auxiliary ex-core neutron detector. To find out the optimal thickness of polyethylene moderator, preliminary simulation and experiments are performed. And sensitivity simulation for detector moderator at actual reactor is performed by DORT code

  11. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  12. A seismic performance and cost comparison of top and bottom supported liquid metal reactor vessels

    International Nuclear Information System (INIS)

    Carlson, T.M.; Kiciman, O.K.; Petrozelli, J.F.

    1989-01-01

    It is the premise of this paper that the revision of a pool LMR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design. Recent capital cost estimates ($/kWe) for U.S. liquid metal reactor (LMR) plant designs reveal that the balance of plant costs could be reduced below that of the balance of plant costs for a comparable light water reactor plant. However, in regions of high seismicity, non-seismically isolated LMR nuclear steam supply system weights are costs per kWe are two to three times the weights and costs of light water reactor nuclear steam supply systems. While all portions of the LMR nuclear steam supply system require examination for potential cost reductions, the focus of this paper is the reactor vessel (RV) block for a large pool plant

  13. Radiogenic Lead with Dominant Content of 208Pb: New Coolant and Neutron Moderator for Innovative Nuclear Facilities

    Directory of Open Access Journals (Sweden)

    A. N. Shmelev

    2011-01-01

    Full Text Available As a rule materials of small atomic weight (light and heavy water, graphite, and so on are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.

  14. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Appleby, E.R.

    1975-08-01

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

  15. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    Science.gov (United States)

    Sibille, Laurent; Dominques, Jesus A.

    2012-01-01

    The maturation of Molten Regolith Electrolysis (MRE) as a viable technology for oxygen and metals production on explored planets relies on the realization of the self-heating mode for the reactor. Joule heat generated during regolith electrolysis creates thermal energy that should be able to maintain the molten phase (similar to electrolytic Hall-Heroult process for aluminum production). Self-heating via Joule heating offers many advantages: (1) The regolith itself is the crucible material, it protects the vessel walls (2) Simplifies the engineering of the reactor (3) Reduces power consumption (no external heating) (4) Extends the longevity of the reactor. Predictive modeling is a tool chosen to perform dimensional analysis of a self-heating reactor: (1) Multiphysics modeling (COMSOL) was selected for Joule heat generation and heat transfer (2) Objective is to identify critical dimensions for first reactor prototype.

  16. First experience with the new solid methane moderator at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Beliakov, A.A.; Shabalin, E.P.; Tretyakov, I.T.

    2001-01-01

    In the 1999 Fall the solid methane moderator (CM) has been installed and tested at full power at the IBR-2 pulsed reactor. Its main features are a beryllium reflector and a light water premoderator. Radiation load on the methane was three times as much as that of IPNS facility, namely, 0.1 W/g. Effects of temperature, operation time, concentration of a hydrogen scavenger, and annealing procedure on both neutron and service performances were studied. Maximum operation time of a newly loaded portion of methane was 4 days. In this time around 30% of methane is transformed into hydrogen, ethane, and high molecular hydrocarbons, and yet no deterioration in cold neutron intensity was detected. Among new knowledge, the most important are two facts observed: two-fold decrease in hydrogen formation rate when methane is poisoned with 2.5% to 5% of ethylene, and low formation rate of solid, inremovable products of radiolysis - (1.5/3)10 -7 g/J, which means that after 10 years of operation the methane chamber will be filled with only 100 g of residue. Gain of factor 20 in cold neutron flux was obtained as compared to the routine grooved light water moderator. Presently, it is the highest among the intense pulsed neutron sources. (author)

  17. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  18. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  19. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  20. Molten salt oxidation of ion-exchange resins doped with toxic metals and radioactive metal surrogates

    International Nuclear Information System (INIS)

    Yang, Hee-Chul; Cho, Yong-Jun; Yoo, Jae-Hyung; Kim, Joon-Hyung; Eun, Hee-Chul

    2005-01-01

    Ion-exchange resins doped with toxic metals and radioactive metal surrogates were test-burned in a bench-scale molten salt oxidation (MSO) reactor system. The purposes of this study are to confirm the destruction performance of the two-stage MSO reactor system for the organic ion-exchange resin and to obtain an understanding of the behavior of the fixed toxic metals and the sulfur in the cationic exchange resins. The destruction of the organics is very efficient in the primary reactor. The primarily destroyed products such as carbon monoxide are completely oxidized in the secondary MSO reactor. The overall collection of the sulfur and metals in the two-stage MSO reactor system appeared to be very efficient. Over 99.5% of all the fixed toxic metals (lead and cadmium) and radioactive metal surrogates (cesium, cobalt, strontium) remained in the MSO reactor bottom. Thermodynamic equilibrium calculations and the XRD patterns of the spent salt samples revealed that the collected metals existed in the form of each of their carbonates or oxides, which are non-volatile species at the MSO system operating conditions. (author)