WorldWideScience

Sample records for cooled fast breeder

  1. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  2. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  3. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  4. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  5. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  6. 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.

  7. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  8. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  9. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  10. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  11. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions Coimbatore, Tamilnadu (India)

    2013-12-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  12. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  13. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  14. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  15. Helium-cooled molten-salt fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

    1984-12-01

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

  16. Safeguards in prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Tomura, Katsuji; Okuda, Yosihisa; Iwamoto, Tomonori [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1994-12-31

    MONJU is the prototype fast breeder reactor in Japan designed to have the electricity output of 280 MWe. Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga site. The loading of the core fuel assemblies to the core have been started since October 1993 and the pre-operational test is undergoing. MONJU uses 198 MOX fuel assemblies as core fuel and 172 DU assemblies as blanket fuel. Assemblies loaded in core and stored in the ex-vessel storage tank (EVST) exist in liquid sodium. These Pu containing fuel assemblies, MOX and irradiated DU, are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area are requested to be verified. The verification of the flows is designed to be made with fuel flow monitors measuring radiations, which can abridge the inspector attendance during the fuel handling. This paper describes the detailed aspects of the fuel transfers in MONJU facility and the verification of them through flow monitors together with the functions of other safeguards equipments. (author).

  17. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  18. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  19. Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

    Science.gov (United States)

    Clief Pattipawaej, Sandro; Su’ud, Zaki

    2017-01-01

    A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

  20. Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H. (UKAEA Information Services Branch, London); Matthews, J.R. (UKAEA Harwell Lab. (UK). Theoretical Physics Div.); Potter, P.E. (UKAEA Harwell Lab. (UK). Chemistry Div.)

    1989-07-01

    The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO{sub 2} based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: Chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified. (orig.).

  1. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    Science.gov (United States)

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  2. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  3. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  4. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  5. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  6. Safeguards in the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Deshimaru, T.; Tomura, K. [Power Reactor and Nuclear Fuels Development Corporation, Ibaraki-ken (Japan)

    1995-12-31

    MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling.

  7. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  8. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  9. Risk-assessment methodology for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K. O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed.

  10. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    Science.gov (United States)

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%.

  11. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

    2016-11-01

    Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

  12. The Case Against the Fast Breeder Reactor: An Anti-Nuclear Establishment View.

    Science.gov (United States)

    Lovins, Amory B.

    1973-01-01

    Environmentalists lobby points out that hazards which may result from mistakes in proposed fast breeder reactor for additional energy can be detrimental for mankind. Such projects must be carefully planned and cautiously executed. (PS)

  13. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

  14. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

  15. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  16. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Science.gov (United States)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  17. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    Science.gov (United States)

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  18. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  19. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  20. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  1. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  2. Comparison of containment systems for large sodium-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seeman, S.E.; Armstrong, G.R.

    1978-04-01

    Three types of containment configurations for large breeder reactors (1000 MWe/2500 MWt) were compared to determine characteristic responses of each type to selected hypothetical accident conditions. The three types of system configurations were the single, double and containment-confinement designs. Two types of postulated hypothetical accidents were applied as a basis for evaluation--the energetic core disassambly accident and the whole core melt-through accident. More than 20 separate cases were analyzed in which design and accident assumptions were varied. Based on this study the containment-confinement configuration is shown to be the most promising for limiting off-site radiation doses for large sodium-cooled breeder reactors given the specific assumptions on the initial accident conditions.

  3. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  4. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)], E-mail: gpk@igcar.gov.in; Pandey, G.K.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2009-06-15

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  5. Performance characterization of geopolymer composites for hot sodium exposed sacrificial layer in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Haneefa, K. Mohammed, E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2013-12-15

    Highlights: • Performance evaluation of geopolymers subjected to hot liquid sodium is performed. • Apart from mechanical properties, micro-analytical techniques are used for material characterization. • The geopolymer composite showed comparatively lesser damage than conventional cement composites. • Geopolymer technology can emerge as a new choice for sacrificial layer in SCFBRs. - Abstract: A sacrificial layer of concrete is used in sodium cooled fast breeder reactors (SCFBRs) to mitigate thermo-chemical effect of accidentally spilled sodium at and above 550 °C on structural concrete. Performance of this layer is governed by thermo-chemical stability of the ingredients of sacrificial layer concrete. Concrete with limestone aggregate is generally used as a sacrificial layer. Conventional cement based systems exhibit instability in hot liquid sodium environment. Geo-polymer composites are well known to perform excellently at elevated temperatures compared to conventional cement systems. This paper discusses performance of such composites subjected to exposure of hot liquid sodium in air. The investigation includes comprehensive evaluation of various geo-polymer composites before any exposure, after heating to 550 °C in air, and after immersing in hot liquid sodium initially heated to 550 °C in air. Results from the current study indicate that hot liquid sodium produces less damage to geopolymer composites than to the existing conventional cement based system. Hence, the geopolymer technology has potential application in mitigating the degrading effects of sodium fires and can emerge as a new choice for sodium exposed sacrificial layer in SCFBRs.

  6. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  7. Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S. [Kernforschungszentrum Karlsruhe GmbH (Germany); Mattas, R. [Argonne National Lab., IL (United States)

    1994-06-01

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R&D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns.

  8. Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Ezato, Koichiro; Seki, Yohji; Yoshikawa, Akira; Tsuru, Daigo; Akiba, Masato [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

  9. Fast Breeder Blanket Facility (FBBF). Annual report, January 31, 1976--December 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K.O. (ed.)

    1978-01-01

    The work performed in the reporting period was primarily concerned with the construction of the Fast Breeder Blanket Facility (FBBF), acquisition of experimental equipment, outlining the experimental program, preanalysis of the initial loading configuration and investigation of the safety of the initial loading and advanced loadings. The detailed physical description of the FBBF, operational procedures and controls, radiation shielding and experimental equipment are presented. The ability of the FBBF to simulate the blanket spectrum of a large fast breeder reactor is illustrated by comparison of spectra. The source axial distribution, reaction rate comparisons, breeding of plutonium and gamma-ray energy deposition rates are also discussed. Some of the safety aspects of the initial loading and advanced loadings are described. Experimental capabilities of the facility are outlined.

  10. Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

    Directory of Open Access Journals (Sweden)

    Sebastian Vehlken

    2014-09-01

    Full Text Available This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Breeder Reactor (LMFBR technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the "German Manhattan Project" not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and critical technological endeavors. In the face of the risks of nuclear technology, German physicist Wolf Häfele thus announced a novel epistemology of "hypotheticality". In a context where traditional experimental engineering strategies became inappropiate, he called for the application of advanced media technologies: Computer Simulations (CS and Systems Analysis (SA generated computerized spaces for the production of knowledge. In the course of the German Fast Breeder program, such methods had a twofold impact. One the one hand, Häfele emphazised – as the "father of the German Fast Breeder" – the utilization of CS for the actual planning and construction of the novel reactor type. On the other, namely as the director of the department of Energy Systems at the International Institute for Applied Systems Analysis (IIASA, Häfele advised SA-based projections of energy consumption. These computerized scenarios provided the rationale for the conception of Fast Breeder programs as viable and necessary alternative energy sources in the first place. By focusing on the role of the involved CS techniques, the paper thus investigates the intertwined systems thinking of nuclear facilities’s planning and construction and the design of large-scale energy consumption and production scenarios in the 1970s and 1980s, as well as their conceptual afterlives in our contemporary era of computer simulation.

  11. Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed Haneefa, K., E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2014-08-15

    Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO{sub 3}, widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone.

  12. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  13. Cellular convection in vertical annuli of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hemanath, M.G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)], E-mail: hemanath@igcar.gov.in; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2007-08-15

    In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.

  14. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  15. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  16. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  17. Fast Breeder Blanket Facility (FBBF). Quarterly progress report, January 1, 1976--August 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K.O. (ed.)

    1976-08-01

    The work performed was primarily concerned with the preparation of the experiments to be performed on the Fast Breeder Blanket Facility (FBBF) and the corresponding analysis. The work on the experimental program has been started. Since experiments are subject to safety constraints, a safety investigation program (for a hypothetically flooded facility) is reported. The neutronics part of the preanalysis is also reported. The testing of the first configuration has largely been prepared. The identification of the experiment need has been worked on extensively, largely through unsponsored research which had been started before the contract became effective. The work done in this area by other groups is being reviewed.

  18. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  19. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  20. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%.

  1. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  2. Development of fuel flow monitoring system in prototype fast breeder reactor 'MONJU'

    Energy Technology Data Exchange (ETDEWEB)

    Tomura, Katsuji; Deshimaru; Takehide; Okuda, Yoshihisa; Ohba, Toshio (Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office); Ishikawa, Kouichi

    1994-06-01

    A new safeguards approach of Prototype Fast Breeder Reactor 'MONJU' has been studied by Japanese Government, IAEA and PNC to meet 1991-1995 safeguards criteria. As the result, a fuel flow monitoring system has been introduced in 'MONJU'. Development of the system has been conducted by PNC and IAEA with technical support of Los Alamos National Laboratory. Safeguards measures in unattended mode with the system can detect fuel loading and unloading into and from the reactor core and distinguish what kind of the fuel. The system are consisted of three monitors using neutron and gamma-ray measurements and video surveillance system. Installation of these monitors was finished by PNC and acceptance test by Japanese Government and IAEA was carried out March, 1992. (author).

  3. Distribution of liquid sodium in the inlet plenum of steam generator in a Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patil, Laxman T. [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Patwardhan, A.W., E-mail: awp@udct.or [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Padmakumar, G.; Vaidyanathan, G. [Experimental Thermal Hydraulics Section, Separation Technology and Hydraulics Division, Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2010-04-15

    Experimental and Computational Fluid Dynamics (CFD) investigations have been carried out on a 1/5th scale model of the inlet plenum of steam generator (SG) used in the Fast Breeder Reactor (FBR) technology. The distribution of liquid sodium in the inlet plenum of the steam generator strongly affects the thermal as well as mechanical performance of the steam generator. In the present work, flow distribution in a scaled down model has been investigated. Various strategies adopted for obtaining uniform flow distribution have been evaluated. Experiments have been conducted to measure the axial and radial velocity distributions using Ultrasonic Velocity Profiler (UVP) under a variety of geometries. Computational Fluid Dynamics (CFD) studies have been carried out for various geometries. On the basis of these experiments and CFD simulations, various flow distribution devices have been compared.

  4. Compendium of computer codes for the safety analysis of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.

  5. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  6. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  7. Boosted Fast Flux Loop Alternative Cooling Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

    2007-08-01

    The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and

  8. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  9. Assessment of gel-sphere-pac fuel for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lackey, W J; Selle, J E [comps.

    1978-10-01

    An assessment of the state of the art for the gel-sphere-pac process was undertaken to provide a sound basis for further development of the technology. Information is provided on sol preparation, sphere forming, drying, sintering, characterization, loading, fuel rod inspection, and irradiation performance. In addition, discussions are included on: evaluation of the potential for scale-up to production capacities, potential problems associated with remote operation, and future work required to further develop the technology. Three techniques are available for microsphere production: (1) internal gelation, (2) external gelation, and (3) gelation by water extraction. Each has its own advantages and disadvantages; for example, internal gelation appears better suited to the preparation of large spheres than the other processes. Numerous advantages and disadvantages are discussed in detail. Scale-up or remote operation of these techniques appears achievable, although some would require less development than others. Techniques have been developed for drying and sintering spheres. Extensive technology has been developed for sphere characterization, handling, and the loading and inspection of fuel pins. Data available to date indicates that sphere-pac oxide fuel will perform similarly to pellet oxide fuels under fast breeder reactor operating conditions. Gel-sphere-pac technology also appears attractive for carbide fuels.

  10. Design of improved thermometer for the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Shimano, Kunio; Ito, Kenji; Tomobe, Katsuma [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Monju Construction Office, Tsuruga, Fukui (Japan)

    2002-12-01

    The thermometer design for the secondary coolant system was improved to prevent recurring failure of the thermometer well due to flow-induced vibration, the direct cause of the sodium leak incident of the prototype fast breeder reactor 'MONJU'. To satisfy the requirements of average temperature measurement, response time (within 20 seconds), avoidance and restraint of synchronized vibration, the insertion length of thermometer wells into the pipe was shortened to 110 mm for the response requirement and 60 mm for the no response requirement with a tapered shape. To simplify the installation, thermometer wells are mounted on the existing nozzles. To confirm the suitability of the design, analyses and experiments using the final design of the improved thermometer were performed. By analytical evaluation of flow-induced vibration and strength, the structural integrity was confirmed. Additionally, through flow-induced vibration experience, analyses of vibration characteristics confirmed the suitability. Furthermore, manufacture and welding of the thermometer wells on the existing nozzles were confirmed to be possible. (author)

  11. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  12. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rajan Babu, V., E-mail: vrb@igcar.gov.i [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India); Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India)

    2010-07-15

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO{sub 2} process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  13. Slow and stunted: Plutonium accounting and the growth of fast breeder reactors in India

    Energy Technology Data Exchange (ETDEWEB)

    Ramana, M.V. [Program in Science, Technology and Environmental Policy and Program on Science and Global Security, Woodrow Wilson School of Public and International Affairs, Princeton University, Princeton, NJ 08542 (United States); Suchitra, J.Y. [Independent Researcher, 303, Meenakshi Apartments, 45 Ranga Rao Road, Shankarapuram, Bangalore 560004 (India)

    2009-12-15

    The Indian Department of Atomic Energy (DAE) has projected a large growth of nuclear power in the country predominantly based on breeder reactors. These projections use a simplistic methodology that does not carefully account for the availability of plutonium that is required to fuel breeder reactors. In this paper, we demonstrate that this methodology is problematic, in particular that it would result in negative balances of plutonium if the DAE's projections were to come true. The DAE's projections also ignore constraints coming from reprocessing capacity in the country. As an alternative, we project the possible growth of nuclear power based on breeder reactors using a methodology consistent with plutonium constraints. The resulting breeder reactor capacity will be only between 17% and 40% of the DAE's projections, and will likely never constitute a major source of electricity in India for several decades at the very least. (author)

  14. REVIEW OF REACTOR SAFETY ANALYSES OF FAST AND LIQUID METAL COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Shaver, R. E.; Wittenbrock, N. G.

    1967-11-01

    Safety analysis reports on United States fast and liquid metal cooled reactors were reviewed to gain a better understanding of the safety philosophy applied to the design of these facilities. This information was compiled to help guide the design and safety analysis of the Fast Flux Test Facility. No attempt was made to draw conclusions concerning the relative merit of different approaches and philosophies used by different reactor design teams. The facilities reviewed were; Enrico Fermi Atomic Power Plant (FERMI) Hallam Nuclear Power Facility (HALLAM) Southwest Experimental Fast Oxide Reactor (SEFOR) Fast Reactor Test Facility (FARET) Experimental Breeder Reactor No. 1 (EBR-I) Experimental Breeder Reactor No. 2 (EBR-II) Fast Reactor Zero Power Experiment (ZPR - III). The information gathered from the safety analysis reports is tabulated under these headings: Control and Safety Systems; Reactor Protection Systems; Backup Systems; Containment or Confinement Systems; Inherent Reactivity Effects and Important Physics Parameters; Fuel and Fuel Handling; Accidents Considered and Chemical Problems; Site; Exhaust Ventilation System; and Waste Effluents.

  15. Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

    Indian Academy of Sciences (India)

    G Pandikumar; A John Arul; P Puthiyavinayagam; P Chellapandi

    2015-10-01

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

  16. Research and development in welding and hardfacing towards construction of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2012-07-01

    India's 500MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction at Kalpakkam and this reactor is expected to be commissioned in the year 2013. Extensive research and development activities in various fields like material development, welding, forming, non-destructive testing etc. were undertaken before the actual construction of the reactor began. Many of these activities are still continuing with the objectives of conducting functional tests, generating data, validating the design and meeting the various regulatory requirements. In welding, initial challenge was to develop indigenous welding consumables with a specification more stringent than that is given in most of the national and international standards. The welding consumable specified for 316LN austenitic stainless steel is E316-15M with strict control on delta ferrite content, toughness requirement after 750 C/100 h ageing to ensure adequate resistance to embrittlement during prolonged high temperature exposure and good slag detachability. This consumable was successfully developed in collaboration with Indian consumable manufacturers and is being used for fabrication of almost all PFBR components and piping made of 316LN stainless steel. Similarly, electrodes of welding of modified 9Cr-1Mo steel (material of construction for PFBR steam generator) with requirement of RTNDT requirement of ≤ -5 C was also developed indigenously. Extensive studies were also carried out on weldability of various austenitic stainless steels and modified 9Cr-1Mo steel used in PFBR. Hot cracking susceptibility of alloy D9 (15Cr-15Ni-2Mo-Ti alloy), the material chosen for fuel clab tube and fuel sub-assembly, and 316LN stainless steel was extensively studied using varestraint testing and Gleeble simulation. Results from these steels were used in developing welding procedures for various reactor components. Hydrogen assisted cracking susceptibility (HAC) of modified 9Cr-1Mo steel was studied using

  17. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    Energy Technology Data Exchange (ETDEWEB)

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  18. Fast cooling techniques for gravitational wave antennas

    CERN Document Server

    Furtado, S R

    2002-01-01

    The resonant-mass technique for the detection of gravitational waves may involve, in the near future, the cooling of very large masses (about 100 tons) from room temperature (300 K) to extreme cryogenic temperatures (20 mK). To cool these detectors to cryogenic temperatures an exchange gas (helium) is used, and the heat is removed from the antenna to the cold reservoir by thermal conduction and natural convection. With the current technique, cooling times of about 1 month can be obtained for cylindrical bar antennas of 2.5 tons. Should this same technique be used to cool a 100 ton spherical antenna the cooling time would be about 10 months, making the operation of these antennas impracticable. In this paper, we study the above-mentioned cooling technique and others, such as thermal switching and forced convection from room temperature to liquid nitrogen temperature (77 K) using an aluminium truncated icosahedron of 19 kg weight and 25 cm diameter.

  19. A vision of inexhaustible energy: The fast breeder reactor in Swedish nuclear power history 1945-80; Visionen om outtoemlig energi: Bridreaktorn i svensk kaernkraftshistoria 1945-80

    Energy Technology Data Exchange (ETDEWEB)

    Fjaestad, Maja, E-mail: majaf@k_th.se

    2010-03-15

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and 1960s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy, and thereby connected it to utopian ideas about an eternal supply of energy, Furthermore. the ideas of breeder reactors were a vital part of the post-war visions about the nuclear future. This dissertation investigates the plans for breeder reactors in Sweden, connecting them to the contemporary development of nuclear power with heavy or light water and the discussions of nuclear weapons, as well as to the general visions of a prosperous technological future. The history of the Swedish breeder reactor is traced from high hopes in the beginning, via the fiasco of the Swedish heavy water program, partly focusing on the activities at the company AB Atomenergi and investigating how it planned and argued for its breeder program and how this was received by the politicians. The story continues into the intensive environmental movement in the 1970s, ending with the Swedish referendum on nuclear energy in 1980, which can be seen as the final point for the Swedish breeder. The thesis discusses how the nuclear breeder reactor was transformed from an argument for nuclear power to an argument against it. The breeder began as a part of the vision of a society with abundant energy, but was later seen as a threat against the new sustainable world. The nuclear breeder reactor is an example of a technological vision that did not meet its industrial expectations. But that does not prevent the fact that breeder was an influential technology in an age where important decisions about nuclear energy were made. The thesis argues that important decisions about the contemporary reactors were taken with the idea that they in a foreseeable future would be replaced with the efficient breeder. And the last word on the breeder reactor is not said - today, reactor engineers around the world are

  20. Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

    2016-02-15

    Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

  1. Manufacturing pre-qualification of a Short Breeder Unit mockup (SHOBU) as part of the roadmap toward the out-of-pile validation of a full scale Helium Cooled Pebble Bed Breeder Unit

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, Francisco A., E-mail: francisco.hernandez@kit.edu [Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology, Karlsruhe (Germany); Rey, Jorg; Neuberger, Heiko [Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology, Karlsruhe (Germany); Krasnorutskyi, Sergii [Institut für Füge- und Schweißtechnik (IFS), Technische Universität Braunschweig, Braunschweig (Germany); Niewöhner, Reinhard [Forschungsgesellschaft Umformtechnik mbH (FGU), Stuttgart (Germany); Felde, Alexander [Institut für Umformtechnik (IFU), Universität Stuttgart, Stuttgart (Germany)

    2015-10-15

    Highlights: • A relevant mockup of a HCPB Breeder Unit for ITER (SHOBU) has been manufactured. • The manufacturing technologies used in SHOBU and its assembly sequence are reported. • Preliminary qualification of the welds has been successfully done after codes. • Future work foreseen to manufacture a feasibility mockup according to RCC-MRx code. - Abstract: The key components of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) in ITER are the Breeder Units (BU). These are the responsible for the tritium breeding and part of the heat extraction in the HCPB TBM. After a detailed design and engineering phase performed during the last years in the Karlsruhe Institute of Technology (KIT), a reference model for the manufacturing of a HCPB BU mock-up has been obtained. The mid-term is the out-of-pile qualification of the thermal and thermo-mechanical performance of a full-scale HCPB BU mock-up in a dedicated helium loop. Several key manufacturing technologies have been developed for the fabrication of the HCPB BU. In order to pre-qualify these techniques, a Short Breeder Unit mock-up (SHOBU) is under construction and to be tested. This paper aims at describing the relevance of SHOBU with a full-scale HCPB BU, the constitutive parts of SHOBU, the manufacturing and joining technologies involved, the assembly sequence (taking into consideration functional steps like its filling with Li{sub 4}SiO{sub 4} pebbles or its assembly in the HCPB TBM) and the welding procedures studied. The paper concludes with a description of the required pre-qualification tests performed to SHOBU, i.e. pressure and leak tightness tests, according to the standards.

  2. Fast cooling of trapped ions using the dynamical Stark shift

    Energy Technology Data Exchange (ETDEWEB)

    Retzker, A [Institute for Mathematical Sciences, Imperial College London, SW7 2PE (United Kingdom); Plenio, M B [Institute for Mathematical Sciences, Imperial College London, SW7 2PE (United Kingdom)

    2007-08-15

    A laser cooling scheme for trapped ions is presented which is based on the fast dynamical Stark shift gate, described in (Jonathan et al 2000 Phys. Rev. A 62 042307). Since this cooling method does not contain an off resonant carrier transition, low final temperatures are achieved even in a traveling wave light field. The proposed method may operate in either pulsed or continuous mode and is also suitable for ion traps using microwave addressing in strong magnetic field gradients.

  3. Longitudinal dynamics of laser-cooled fast ion beams

    DEFF Research Database (Denmark)

    Weidemüller, M.; Eike, B.; Eisenbarth, U.

    1999-01-01

    We present recent results of our experiments on laser cooling of fast stored ion beams at the Heidelberg Test Storage Ring. The longitudinal motion of the ions is directly cooled by the light pressure force, whereas efficient transverse cooling is obtained indirectly by longitudinal......-transverse coupling mechanisms. Laser cooling in novel bunch forms consisting of square-well buckets leads to longitudinally space-charge dominated beams. The observed longitudinal ion density distributions can be well described by a self-consistent mean-field model based on a thermodynamic Debye-Huckel approach....... When applying laser cooling in square-well buckets over long time intervals, hard Coulomb collisions suddenly disappear and the longitudinal temperature drops by about a factor of three. The observed longitudinal behaviour of the beam shows strong resemblance with the transition to an Coulomb...

  4. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  5. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  6. Linearized model for the hydrodynamic stability investigation of molten fuel jets into the coolant of a Liquid Metal Fast Breeder Reactor (LMFBR)

    Science.gov (United States)

    Hartel, K.

    1986-02-01

    The hydrodynamic stability of liquid jets in a liquid continuum, both characterized by low viscosity was analyzed. A linearized mathematical model was developed. This model enables the length necessary for fragmentation of a vertical, symmetric jet of molten fuel by hydraulic forces in the coolant of a liquid metal fast breeder reactor to be evaluated. On the basis of this model the FRAG code for numerical calculation of the hydrodynamic fragmentation mechanism was developed.

  7. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  8. Kinematic dynamo action in a network of screw motions; application to the core of a fast breeder reactor

    Science.gov (United States)

    Plunian, F.; Marty, P.; Alemany, A.

    1999-03-01

    Most of the studies concerning the dynamo effect are motivated by astrophysical and geophysical applications. The dynamo effect is also the subject of some experimental studies in fast breeder reactors (FBR) for they contain liquid sodium in motion with magnetic Reynolds numbers larger than unity. In this paper, we are concerned with the flow of sodium inside the core of an FBR, characterized by a strong helicity. The sodium in the core flows through a network of vertical cylinders. In each cylinder assembly, the flow can be approximated by a smooth upwards helical motion with no-slip conditions at the boundary. As the core contains a large number of assemblies, the global flow is considered to be two-dimensionally periodic. We investigate the self-excitation of a two-dimensionally periodic magnetic field using an instability analysis of the induction equation which leads to an eigenvalue problem. Advantage is taken of the flow symmetries to reduce the size of the problem. The growth rate of the magnetic field is found as a function of the flow pitch, the magnetic Reynolds number (Rm) and the vertical magnetic wavenumber (k). An [alpha]-effect is shown to operate for moderate values of Rm, supporting a mean magnetic field. The large-Rm limit is investigated numerically. It is found that [alpha]=O(Rm[minus sign]2/3), which can be explained through appropriate dynamo mechanisms. Either a smooth Ponomarenko or a Roberts type of dynamo is operating in each periodic cell, depending on k. The standard power regime of an industrial FPBR is found to be subcritical.

  9. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  10. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  11. Preliminary Assessment of a Debris Bed Cooling Performance for Demonstration Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chung Ho; Park, Chang Gyu; Song, Hoon; Kim, Young Gyun; Jeong, Hae Yong; Chang, Jin Wook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In the case of the sodium-cooled fast reactor such as KALIMER-600, Hypothetical Core Disruptive Accident (HCDA) attributed from mass nuclear fuel melting is unlikely to occur due to defense in depth concepts to meet requirements of redundancy and diversity. Multiple faults such as loss of flow, loss of heat sink, or transient overpower without scram are to lead rising the power level until cladding failure as reactivity increasing. The fact that metallic fuel melts at a lower temperature than the cladding allows significant in-pin- fuel motion to occur prior to cladding failure. Also, the combination of Doppler and axial expansion feedback and negative feedback associated with the in-pin fuel relocation prevents the reactivity from reaching prompt critical. Finally, the resulting reactivity and power reductions help prevent fuel temperatures from rising more than the fuel melting temperature. It is more difficult to occur HCDA in a metallic fueled core because reactor power and heat removal capability is maintained in balance by inherent safety characteristics However, for the future design of sodium-cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth considering due to the triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. Accordingly, evaluation of a packed debris bed cooling performance with single phase flow for demonstration sodium-cooled fast reactor was carried out for proof of the in-vessel retention of the core debris

  12. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  13. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  14. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

  15. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  16. A method for improvement of safety features of large fast breeder reactors. Numerical simulation of unprotected loss-of-flow accident in an LMFBR equipped with gas-expansion modules

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Masayoshi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Murakami, Tomoko; Kawashima, Katsuyuki; Watari, Yoshio; Nakao, Noboru; Miura, Masanori

    1995-04-01

    Numerical simulation of an unprotected loss-of-flow (ULOF) accident has been performed for a large liquid-metal-cooled fast breeder reactor (LMFBR) equipped with gas expansion modules (GEMs) in the radial periphery of the reactor core. The effectiveness of the GEMs in small fast reactors was demonstrated already in the passive safety testing in the Fast Flux Test Facility. According to neutronic calculations based on the transport theory, even in large reactors of electrical power 600 to 1,300 MW, the reactivity worth of GEMs, which replace one layer of radial blanket fuel subassemblies, ranges from -1.9$ to -1.4$, depending on the size of the core. A simulation of ULOF transient was performed with a 5.5s flow-halving time in a 600 MWe LMFBR equipped with GEMs of -1.9$ reactivity worth. The result showed that, if 10% of the rated core coolant flow by pony motors was available following the main pump coastdown, the GEM reactivity alone could bring the reactor subcritical and the predicted maximum coolant temperature was substantially lower than the sodium boiling point. The reactivity worth calculations, a modeling of gas expansion behavior, and ULOF simulation together with needs of further development for the GEM application are described. (author).

  17. Localized corrosion studies on materials proposed for a safety-grade sodium-to- air decay-heat removal system for fast breeder reactors

    Science.gov (United States)

    Kamachi Mudali, U.; Khatak, H. S.; Dayal, R. K.; Gnanamoorthy, J. B.

    1993-02-01

    The present investigation was carried out to assess the localized corrosion resistance of materials proposed for the construction of the safety-grade sodium-to-air decay-heat removal system for fast breeder reactors. The materials, such as Alloy 800,9Cr-lMo steel, and type 316LN stainless steel, in different microstructural conditions were assessed for pitting and stress-corrosion cracking resistances in a chloride medium. The results indicated that 9Cr-lMo steel in the normalized and tempered condition can be considered for the above application from the standpoint of corrosion resistance.

  18. High performance infrared fast cooled detectors for missile applications

    Science.gov (United States)

    Reibel, Yann; Espuno, Laurent; Taalat, Rachid; Sultan, Ahmad; Cassaigne, Pierre; Matallah, Noura

    2016-05-01

    SOFRADIR was selected in the late 90's for the production of 320×256 MW detectors for major European missile programs. This experience has established our company as a key player in the field of missile programs. SOFRADIR has since developed a vast portfolio of lightweight, compact and high performance JT-based solutions for missiles. ALTAN is a 384x288 Mid Wave infrared detector with 15μm pixel pitch, and is offered in a miniature ultra-fast Joule- Thomson cooled Dewar. Since Sofradir offers both Indium Antimonide (InSb) and Mercury Cadmium Telluride technologies (MCT), we are able to deliver the detectors best suited to customers' needs. In this paper we are discussing different figures of merit for very compact and innovative JT-cooled detectors and are highlighting the challenges for infrared detection technologies.

  19. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    Science.gov (United States)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  20. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  1. Experimental Progress in Fast Cooling in the ESR

    CERN Document Server

    Steck, Markus; Beller, Peter; Franzke, Bernhard; Nolden, Fritz

    2005-01-01

    The ESR storage ring at GSI is operated with highly charged heavy ions. Due to the high electric charge the ions interact much stronger with electromagnetic fields. Therefore both cooling methods which are applied to stored ions in the ESR, stochastic cooling and electron cooling, are more powerful than for singly charged particles. The experimental results exhibit cooling times for stochastic cooling of a few seconds. For cold ion beams, electron cooling provides cooling times which are one to two orders of magnitude smaller. The beams are cooled to beam parameters which are limited by intrabeam scattering. At small ion numbers, however, intrabeam scattering is suppressed by electron cooling, clear evidence was found that the ion beam forms a one-dimensional ordered structure, a linear chain of ions. The strengths of stochastic cooling and electron cooling are complementary and can be combined favorably. Stochastic cooling is employed for pre-cooling of hot secondary beams followed by electron cooling to pro...

  2. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  3. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  4. Breeder Reactors, Understanding the Atom Series.

    Science.gov (United States)

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  5. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  6. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  7. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  8. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  9. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  10. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    Science.gov (United States)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  11. Numerical studies on the heat transfer and friction characteristics of the first wall inserted with the screw blade for water cooled ceramic breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230037 (China); Ma, Xuebin; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-03-15

    Highlights: • Enhanced heat transfer and friction characteristics of the FW inserted with screw blade is investigated. • The screw blade structure optimization was done on the screw pitch and diameter. • Decreasing screw pitch and increasing screw diameter could further enhance heat transfer accompanied with increasing flow resistance. • Evaluate the overall enhanced heat performance by using the PEC value. - Abstract: The Water Cooled Ceramic Breeder (WCCB) blanket based on Pressurized Water Reactor (PWR) condition is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The first wall (FW) which plays an important part in the blanket design must remove the high heat flux radiated from plasma and nuclear heat deposition on the structure in any operating conditions. In this paper, the characteristics of enhanced heat transfer and friction for the FW with the inserted screw blade are studied by the numerical method. After the comparison between the numerical and experimental results, the standard k–ε turbulent model is selected to do the numerical calculation. The numerical results show that the peak temperature of RAFM steel could be reduced by decreasing screw pitch or increasing screw diameter, while accompanied with ascending flow resistance. Besides, among all of the chosen calculation cases compared with the smooth channel, the maximum value of temperature reduction is 10 °C under the conditions of heat flux of 0.5 MW/m{sup 2} as well as screw pitch of 18 mm and screw diameter of 6 mm. The maximum increment ratio of the friction factor is 257% under the conditions of screw pitch of 10 mm and screw diameter of 4 mm. Furthermore, screw blade of 74 mm pitch and 4 mm diameter presents the highest overall performance evaluation criterion (PEC) value of 0.93 under Reynolds number of 270 000 conditions, and shows the best overall heat transfer enhancement performance.

  12. Fast optical cooling of a nanomechanical cantilever by a dynamical Stark-shift gate

    Science.gov (United States)

    Yan, Leilei; Zhang, Jian-Qi; Zhang, Shuo; Feng, Mang

    2015-10-01

    The efficient cooling of nanomechanical resonators is essential to exploration of quantum properties of the macroscopic or mesoscopic systems. We propose such a laser-cooling scheme for a nanomechanical cantilever, which works even for the low-frequency mechanical mode and under weak cooling lasers. The cantilever is coupled by a diamond nitrogen-vacancy center under a strong magnetic field gradient and the cooling is assisted by a dynamical Stark-shift gate. Our scheme can effectively enhance the desired cooling efficiency by avoiding the off-resonant and undesired carrier transitions, and thereby cool the cantilever down to the vicinity of the vibrational ground state in a fast fashion.

  13. Fast optical cooling of a nanomechanical cantilever by a dynamical Stark-shift gate

    CERN Document Server

    Yan, Leilei; Zhang, Shuo; Feng, Mang

    2014-01-01

    The efficient cooling of the nanomechanical resonators is essential to exploration of quantum properties of the macroscopic or mesoscopic systems. We propose such a laser-cooling scheme for a nanomechanical cantilever, which works even for the low-frequency mechanical mode and under weak cooling lasers. The cantilever is attached by a diamond nitrogen-vacancy center under a strong magnetic field gradient and the cooling is assisted by a dynamical Stark-shift gate. Our scheme can effectively enhance the desired cooling efficiency by avoiding the off-resonant and unexpected carrier transitions, and thereby cool the cantilever down to the vicinity of the vibrational ground state in a fast fashion.

  14. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyasheela, T., E-mail: sheela@igcar.gov.in; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.

    2015-09-15

    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  15. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  16. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    OpenAIRE

    Pengcheng Zhao; Kangli Shi; Shuzhou Li; Jingchao Feng; Hongli Chen

    2016-01-01

    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  17. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Science.gov (United States)

    Ioltukhovskiy, A. G.; Leonteva-Smirnova, M. V.; Solonin, M. I.; Chernov, V. M.; Golovanov, V. N.; Shamardin, V. K.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  18. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ioltukhovskiy, A.G. E-mail: iral@bochvar.ru; Leonteva-Smirnova, M.V.; Solonin, M.I.; Chernov, V.M.; Golovanov, V.N.; Shamardin, V.K.; Bulanova, T.M.; Povstyanko, A.V.; Fedoseev, A.E

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a {delta}-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 deg. C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 deg. C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  19. LMFBR (LIQUID METAL FAST BREEDER REACTOR) READTION RATE AND DOSIMETRY 3RD QUARTERLY PROGRESS REPORT DECEMBER 1971 JANUARY FEBRUARY 1972

    Energy Technology Data Exchange (ETDEWEB)

    MCELROY WN

    1972-03-01

    This report was compiled at the Hanford Engineering Development Laboratory operated by Westinghouse Hanford Company, a subsidiary of Westinghouse Electric Corporation, for the United States Atomic Energy Commission, Division of Reactor Development and Technology, under Contract No. AT (45-1) 2170. It describes technical progress made in the Interlaboratory LMFBR Reaction Rate Program during the reporting period. The Interlaboratory LMFBR Reaction Rate (ILRR) program has been established by USAEC/RDT to develop a capability to accurately measure neutron-induced reaction rates for LMFBR fuels and materials development programs. The initial goal for the principal fission reactions, {sup 235}U, {sup 238}U, and {sup 239}Pu, is an accuracy to within {+-}5 at the 95% confidence level. Accurate measurement of other fission and non-fission reactions will be required, but to a lesser accuracy, between {+-}5 to 10% at the 95% confidence level. A secondary program objective is improvement in knowledge of the nuclear parameters involved in fuels and materials dosimetry measurements of neutron flux, spectra, fluence, and burnup. These accuracy goals for the ILRR program are severe; measurements of fast-neutron-induced reaction rates have not been rapidly moving toward this level of precision. Using a number of techniques in well established neutron environments of current interest for fast reactor development and critically evaluating the results will help establish existing levels of accuracy and indicate the scale of effort required for improvement. To accomplish the objectives of this program, reliable and documented experimental values of reaction rates and ratios will be determined for various well established and permanent neutron fields. The Coupled Fast Reactivity Measurement Facility (CFRMF) at Aerojet Nuclear Company (ANC) is the first neutron field being studied because of the similarity of its spectrum to that of a fast reactor and the range and reproducibility of

  20. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  1. SIEX: a correlated code for the prediction of Liquid Metal Fast Breeder Reactor (LMFBR) fuel thermal performance

    Energy Technology Data Exchange (ETDEWEB)

    Dutt, D.S.; Baker, R.B.

    1974-01-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperatures. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ``consistant`` set of material properties and correlated models provided by the code. The development of physically based correlations to model certain of the phenomana has resulted in a computer program which provides reliable estimates of thermal performance characteristics, yet requires a small amount of core storage and computer running time.

  2. Fast Quasi-Adiabatic Gas Cooling: An Experiment Revisited

    Science.gov (United States)

    Oss, S.; Gratton, L. M.; Calza, G.; Lopez-Arias, T.

    2012-01-01

    The well-known experiment of the rapid expansion and cooling of the air contained in a bottle is performed with a rapidly responsive, yet very cheap thermometer. The adiabatic, low temperature limit is approached quite closely and measured with our apparatus. A straightforward theoretical model for this process is also presented and discussed.…

  3. Case-Based Reasoning(CBR) Model for Ultra-Fast Cooling in Plate Mill

    Institute of Scientific and Technical Information of China (English)

    HU Xiao,WANG Zhaodong,; WANG Guodong

    2014-01-01

    New generation thermo-mechanical control process(TMCP) based on ultra-fast cooling is being widely adopted in plate mill to product high-performance steel material at low cost. Ultra-fast cooling system is complex because of optimizing the temperature control error generated by heat transfer mathematical model and process parameters. In order to simplify the system and improve the temperature control precision in ultra-fast cooling process, several existing models of case-based reasoning(CBR) model are reviewed. Combining with ultra-fast cooling process, a developed R5 CBR model is proposed, which mainly improves the case representation, similarity relation and retrieval module. Certainty factor is defined in semantics memory unit of plate case which provides not only internal data reliability but also product performance reliability. Similarity relation is improved by defined power index similarity membership function. Retrieval process is simplified and retrieval efficiency is improved apparently by windmill retrieval algorithm. The proposed CBR model is used for predicting the case of cooling strategy and its capability is superior to traditional process model. In order to perform comprehensive investigations on ultra-fast cooling process, different steel plates are considered for the experiment. The validation experiment and industrial production of proposed CBR model are carried out, which demonstrated that finish cooling temperature(FCT) error is controlled within±25℃ and quality rate of product is more than 97%. The proposed CBR model can simplify ultra-fast cooling system and give quality performance for steel product.

  4. Case-based reasoning(CBR) model for ultra-fast cooling in plate mill

    Science.gov (United States)

    Hu, Xiao; Wang, Zhaodong; Wang, Guodong

    2014-11-01

    New generation thermo-mechanical control process(TMCP) based on ultra-fast cooling is being widely adopted in plate mill to product high-performance steel material at low cost. Ultra-fast cooling system is complex because of optimizing the temperature control error generated by heat transfer mathematical model and process parameters. In order to simplify the system and improve the temperature control precision in ultra-fast cooling process, several existing models of case-based reasoning(CBR) model are reviewed. Combining with ultra-fast cooling process, a developed R5 CBR model is proposed, which mainly improves the case representation, similarity relation and retrieval module. Certainty factor is defined in semantics memory unit of plate case which provides not only internal data reliability but also product performance reliability. Similarity relation is improved by defined power index similarity membership function. Retrieval process is simplified and retrieval efficiency is improved apparently by windmill retrieval algorithm. The proposed CBR model is used for predicting the case of cooling strategy and its capability is superior to traditional process model. In order to perform comprehensive investigations on ultra-fast cooling process, different steel plates are considered for the experiment. The validation experiment and industrial production of proposed CBR model are carried out, which demonstrated that finish cooling temperature(FCT) error is controlled within ±25°C and quality rate of product is more than 97%. The proposed CBR model can simplify ultra-fast cooling system and give quality performance for steel product.

  5. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  6. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) heat transport system dynamics and steam generator control: Figures

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Dynamic modeling of LMFBR heat transport system is discussed. Uncontrolled transient behavior of individual components and of the integrated heat transport system are considered. For each component, results showing specific dynamic features of the component and/or model capability were generated. Controlled dynamic behavior for alternative steam generator control systems during forced and natural sodium coolant circulation was analyzed. Combined free and forced convection of laminar and turbulent vertical pipe flow of liquid metals was investigated.

  7. 快堆钠回路水锤程序开发与应用%Waterhammer Program Development and Application for Fast Breeder Reactor's Sodium Circus

    Institute of Scientific and Technical Information of China (English)

    文静; 栾霖; 金德圭; 陆道纲; 汤荣铭

    2001-01-01

    研究开发了快堆钠回路水锤分析专用程序WHA。该程序在一维特征线法(MOC)传统的压力波传播数学模型中补充了钠腔-气腔外边界模型,并采用气泡离散模型模拟低压液柱分离中的蒸汽穴的生成与溃灭。程序用FORTRAN90语言对快堆实验钠回路ESPRESSO中由于阀门的快速开启与关闭引起的压力波传播进行了分析计算。计算结果表明:将钠腔-气腔引入水锤压力波传播的数学模型进行程序计算的结果是合理的。%Based on one-dimensional method of characteristics(MOC), anumerical model of pressure-wave progation is presented in the paper. A special code is programmed to analyze and calculate waterhammer resulted from rapid opening or closing of valve in the experimental sodium circus of fast breeder reactor(FBR). In the model, a new outer boundary condition, sodium-cavity is included. Model of bubble's discrete distribution is adopted to simulate generation and collapse of the bubble with the pressure's decreasing and increasing. The results demonstrate that the model of pressure-wave progation is valid.

  8. Cooling rate dependence of undercooling of pure Sn single drop by fast scanning calorimetry

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Bin [Shanghai University, Shanghai Key Laboratory of Modern Metallurgy and Materials Processing, Shanghai (China); University of Rostock, Institute of Physics, Rostock (Germany); Gao, Yulai; Zou, Changdong; Zhai, Qijie [Shanghai University, Shanghai Key Laboratory of Modern Metallurgy and Materials Processing, Shanghai (China); Abyzov, A.S. [Kharkov Institute of Physics and Technology, National Science Center, Kharkov (Ukraine); Zhuravlev, E.; Schmelzer, J.W.P.; Schick, C. [University of Rostock, Institute of Physics, Rostock (Germany)

    2011-07-15

    Non-adiabatic fast scanning calorimetry has been developed to in-situ measure the response of single metallic drops to temperature changes in a large range of cooling rate spanning four orders of magnitude. In particular, the effect of cooling rate on the degree of undercooling of one 10 {mu}m Sn-drop is studied. The experimental results show that the undercooling could be increased first significantly with increasing cooling rate going over to a stage of slow increase for high cooling rates, which indicates a shelf-like dependence of undercooling level on cooling rate before and after a ''crossover'' at a cooling rate of about 1000 K/s where two different heterogeneous mechanisms act simultaneously. First theoretical estimates are developed on the specific feature of the heterogeneous nucleation process of the effect analyzed and possible directions of further research are anticipated. (orig.)

  9. Fast cooling of trapped ions using the dynamical Stark shift gate

    OpenAIRE

    Retzker, A.; Plenio, M. B.

    2006-01-01

    A laser cooling scheme for trapped ions is presented which is based on the fast dynamical Stark shift gate, described in [Jonathan etal, PRA 62, 042307]. Since this cooling method does not contain an off resonant carrier transition, low final temperatures are achieved even in traveling wave light field. The proposed method may operate in either pulsed or continuous mode and is also suitable for ion traps using microwave addressing in strong magnetic field gradients.

  10. Hot Rolled Strip Re-reddening Temperature Changing Law during Ultra-fast Cooling

    Institute of Scientific and Technical Information of China (English)

    Lian-yun JIANG; Chun-jiang ZHAO; Jian-hui SHI; Guo YUAN; Xue-qiang WANG; Qing-xue HUANG

    2015-01-01

    Temperature deviation between surface and the center of hot rolled strip is formed during ultra-fast cooling (UFC). Surface temperature would rise when temperature deviation goes up to an extent, and strip re-reddening phenomenon will appear. Strip re-reddening affects the stability of strip microstructure, property and temperature control precision. Thus, it is necessary to conduct research on re-reddening temperature changing law to improve strip property and temperature control precision. Strip temperature trends for various strip thicknesses and ultra-fast cooling rates were obtained by numerical calculation method. Re-reddening temperature, temperature deviation between surface and center, and boundary layer position changing law were obtained. By comparison, some conclusions were obtained: UFC re-reddening temperature and laminar cooling (LC) re-reddening temperature were linear to ultra-fast cooling rate respectively. Ultra-fast cooling rate affected UFC re-reddening temperature greatly, but it had little effect on LC re-reddening temperature. Equations which were used to calculate UFC re-reddening temperature, LC re-reddening temperature and maximum temperature deviation were obtained. The position of boundary layer stayed in 1/4 strip thickness.

  11. Fast cooling of bunches in compton storage rings*

    CERN Document Server

    Bulyak, E; Zimmermann, F

    2011-01-01

    We propose an enhancement of laser radiative cooling by utilizing laser pulses of small spatial and temporal dimensions, which interact only with a fraction of an electron bunch circulating in a storage ring. We studied the dynamics of such electron bunch when laser photons scatter off the electrons at a collision point placed in a section with nonzero dispersion. In this case of ‘asymmetric cooling’, the stationary energy spread is much smaller than under conditions of regular scattering where the laser spot size is larger than the electron beam; and the synchrotron oscillations are damped faster. Coherent oscillations of large amplitude may be damped within one synchrotron period, so that this method can support the rapid successive injection of many bunches in longitudinal phase space for stacking purposes. Results of extensive simulations are presented for the performance optimization of Compton gamma-ray sources and damping rings.

  12. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  13. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  14. Microstructure and Mechanism of Strengthening of Microalloyed Pipeline Steel: Ultra-Fast Cooling (UFC) Versus Laminar Cooling (LC)

    Science.gov (United States)

    Zhao, J.; Wang, X.; Hu, W.; Kang, J.; Yuan, G.; Di, H.; Misra, R. D. K.

    2016-06-01

    A novel thermo-mechanical controlled processing (TMCP) schedule involving ultra-fast cooling (UFC) technique was used to process X70 (420 MPa) microalloyed pipeline steel with high strength-high toughness combination. A relative comparison is made between microstructure and mechanical properties between conventionally processed (CP) and ultra-fast cooled (UFC) pipeline steels, together with differences in strengthening mechanisms with respect to both types of processes. UFC-processed steel exhibited best combination of strength and good toughness compared to the CP process. The microstructure of CP pipeline steel mainly consisted of acicular ferrite (AF), bainitic ferrite (BF), and dispersed secondary martensite/austenite (M/A) constituent and a small fraction of fine quasi-polygonal ferrite. In contrast, the microstructure of UFC-processed pipeline steel was predominantly composed of finer AF, BF, and dispersed M/A constituent. The primary strengthening mechanisms in UFC pipeline steel were grain size strengthening and dislocation strengthening with strength increment of ~277 and ~151 MPa, respectively. However, the strengthening contribution in CP steel was related to grain size strengthening, dislocation strengthening, and precipitation strengthening, and the corresponding strength increments were ~212, ~149 and ~86 MPa, respectively. The decrease in strength induced by reducing Nb and Cr in UFC pipeline steel was compensated by enhancing the contribution of grain size strengthening in the UFC process. In conclusion, cooling schedule of UFC combined with LC is a promising method for processing low-cost pipeline steels.

  15. An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

    2015-02-15

    Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

  16. Fast cooling for a system of stochastic oscillators

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yongxin, E-mail: chen2468@umn.edu; Georgiou, Tryphon T., E-mail: tryphon@umn.edu [Department of Electrical and Computer Engineering, University of Minnesota, 200 Union Street S.E., Minneapolis, Minnesota 55455 (United States); Pavon, Michele, E-mail: pavon@math.unipd.it [Dipartimento di Matematica, Università di Padova, Via Trieste 63, 35121 Padova (Italy)

    2015-11-15

    We study feedback control of coupled nonlinear stochastic oscillators in a force field. We first consider the problem of asymptotically driving the system to a desired steady state corresponding to reduced thermal noise. Among the feedback controls achieving the desired asymptotic transfer, we find that the most efficient one from an energy point of view is characterized by time-reversibility. We also extend the theory of Schrödinger bridges to this model, thereby steering the system in finite time and with minimum effort to a target steady-state distribution. The system can then be maintained in this state through the optimal steady-state feedback control. The solution, in the finite-horizon case, involves a space-time harmonic function φ, and −logφ plays the role of an artificial, time-varying potential in which the desired evolution occurs. This framework appears extremely general and flexible and can be viewed as a considerable generalization of existing active control strategies such as macromolecular cooling. In the case of a quadratic potential, the results assume a form particularly attractive from the algorithmic viewpoint as the optimal control can be computed via deterministic matricial differential equations. An example involving inertial particles illustrates both transient and steady state optimal feedback control.

  17. Fast cooling for a system of stochastic oscillators

    Science.gov (United States)

    Chen, Yongxin; Georgiou, Tryphon T.; Pavon, Michele

    2015-11-01

    We study feedback control of coupled nonlinear stochastic oscillators in a force field. We first consider the problem of asymptotically driving the system to a desired steady state corresponding to reduced thermal noise. Among the feedback controls achieving the desired asymptotic transfer, we find that the most efficient one from an energy point of view is characterized by time-reversibility. We also extend the theory of Schrödinger bridges to this model, thereby steering the system in finite time and with minimum effort to a target steady-state distribution. The system can then be maintained in this state through the optimal steady-state feedback control. The solution, in the finite-horizon case, involves a space-time harmonic function φ, and -logφ plays the role of an artificial, time-varying potential in which the desired evolution occurs. This framework appears extremely general and flexible and can be viewed as a considerable generalization of existing active control strategies such as macromolecular cooling. In the case of a quadratic potential, the results assume a form particularly attractive from the algorithmic viewpoint as the optimal control can be computed via deterministic matricial differential equations. An example involving inertial particles illustrates both transient and steady state optimal feedback control.

  18. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    Science.gov (United States)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  19. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  20. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  1. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all

  2. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  3. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  4. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  5. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  6. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  7. CASY: a dynamic simulation of the gas-cooled fast breeder reactor core auxiliary cooling system. Volume II. Example computer run

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    A listing of a CASY computer run is presented. It was initiated from a demand terminal and, therefore, contains the identification ST0952. This run also contains an INDEX listing of the subroutine UPDATE. The run includes a simulated scram transient at 30 seconds.

  8. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

  9. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  10. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  11. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indic...

  12. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  13. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Science.gov (United States)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  14. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  15. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  16. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  17. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  18. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Darnowski Piotr

    2015-03-01

    Full Text Available An analysis of the influence of addition of minor actinides (MA to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

  19. Numerical studies of fast ion slowing down rates in cool magnetized plasma using LSP

    Science.gov (United States)

    Evans, Eugene S.; Kolmes, Elijah; Cohen, Samuel A.; Rognlien, Tom; Cohen, Bruce; Meier, Eric; Welch, Dale R.

    2016-10-01

    In MFE devices, rapid transport of fusion products from the core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. The first-orbit trajectories of most fusion products from small field-reversed configuration (FRC) devices will traverse the SOL, allowing those particles to deposit their energy in the SOL and be exhausted along the open field lines. Thus, the fast ion slowing-down time should affect the energy balance of an FRC reactor and its neutron emissions. However, the dynamics of fast ion energy loss processes under the conditions expected in the FRC SOL (with ρe fast ions in a cool plasma. As we use explicit algorithms, these simulations must spatially resolve both ρe and λDe, as well as temporally resolve both Ωe and ωpe, increasing computation time. Scaling studies of the fast ion charge (Z) and background plasma density are in good agreement with unmagnetized slowing down theory. Notably, Z-scaling represents a viable way to dramatically reduce the required CPU time for each simulation. This work was supported, in part, by DOE Contract Number DE-AC02-09CH11466.

  20. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  1. LSP simulations of fast ions slowing down in cool magnetized plasma

    Science.gov (United States)

    Evans, Eugene S.; Cohen, Samuel A.

    2015-11-01

    In MFE devices, rapid transport of fusion products, e.g., tritons and alpha particles, from the plasma core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. Through these two processes in the SOL, the fast particle slowing-down time will have a major effect on the energy balance of a fusion reactor and its neutron emissions, topics of great importance. In small field-reversed configuration (FRC) devices, the first-orbit trajectories of most fusion products will traverse the SOL, potentially allowing those particles to deposit their energy in the SOL and eventually be exhausted along the open field lines. However, the dynamics of the fast-ion energy loss processes under conditions expected in the FRC SOL, where the Debye length is greater than the electron gyroradius, are not fully understood. What modifications to the classical slowing down rate are necessary? Will instabilities accelerate the energy loss? We use LSP, a 3D PIC code, to examine the effects of SOL plasma parameters (density, temperature and background magnetic field strength) on the slowing down time of fast ions in a cool plasma with parameters similar to those expected in the SOL of small FRC reactors. This work supported by DOE contract DE-AC02-09CH11466.

  2. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  3. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  4. Interim status report on lead-cooled fast reactor (LFR) research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  5. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  6. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  7. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Science.gov (United States)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  8. Development of objective provision trees for Sodium-Cooled Fast Reactor Defense-in-depth evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    KALIMER is one of sodium-cooled fast reactor and being developed by Korea Atomic Energy Research Institute (KAERI), was developed and suggested in this paper. Developed OPT is for the defense-in-depth level 3, core heat removal safety function. Using OPT method, the evaluation of defense-in-depth implementation for the design features of KALIMER reactors were tried in this study. To utilize the design information of KALIMER, challenges in OPTs which are under development in this study, were identified based on the system physical boundaries. This approach make the identification of possible and postulated challenges much clear and this will be a benefit to further identification of provisions in KALIMER design. OPTs for other levels of defense-in-depth and other safety functions are under development.

  9. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  10. Impact of nuclear data on sodium-cooled fast reactor calculations

    Science.gov (United States)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  11. Impact of nuclear data on sodium-cooled fast reactor calculations

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2016-01-01

    Full Text Available Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  12. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-15

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  13. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kurisaka, K., E-mail: kurisaka.kennichi@jaea.go.jp [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Sakai, T.; Yamano, H. [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Fujita, S.; Minagawa, K. [Department of Mechanical Engineering, School of Engineering, Tokyo Denki University, Tokyo (Japan); Yamaguchi, A.; Takata, T. [Department of Energy and Environment Engineering, Osaka University, Osaka (Japan)

    2014-04-01

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

  14. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

  15. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  16. Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook

    Directory of Open Access Journals (Sweden)

    W. F. G. van Rooijen

    2009-01-01

    Full Text Available A review is given of developments in the area of Gas-Cooled Fast Reactors (GCFR in the period from roughly 1960 until 1980. During that period, the GCFR concept was expected to increase the breeding gain, the thermal efficiency of a nuclear power plant, and alleviate some of the problems associated with liquid metal coolants. During this period, the GCFR concept was found to be more challenging than liquid-metal-cooled reactors, and none were ever constructed. In the second part of the paper, we provide an overview of the investigations on GCFR since the year 2000, when the Generation IV Initiative rekindled interest in this reactor type. The new GCFR concepts focus primarily on sustainable nuclear power, with very efficient resource use, minimum waste, and a very strong focus on (passive safety. An overview is presented of the main design characteristics of these Gen IV GCFRs, and a literature list is provided to guide the interested reader towards more detailed publications.

  17. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  18. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  19. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  20. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  1. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  2. Neutronic assessment of liquid-metal cooled fast reactors using thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pilarski, Stevan [Electricite de France R et D, 1 Avenue du General de Gaulle, 92141 Clamart (France); Institut de Physique Nucleaire d' Orsay, 15 rue Georges Clemenceau 91406 Orsay (France)

    2009-06-15

    The long-term sustainability of atomic fission energy will require the development of new types of reactors, able to exceed the limits of the existing ones in terms of optimal use of natural resources, which clearly necessitates breeding of fissile material. In this context, fast reactors using uranium-plutonium fuel are the most mature solution from an industrial viewpoint. In addition to the obvious interest in terms of fuel resources, there is a major incentive to consider the use of the {sup 232}Th- {sup 233}U fuel cycle as an alternative to the traditional {sup 238}U-{sup 239}Pu cycle for fast reactors: it is an effective way of addressing the safety issue of the highly positive void reactivity effect, which is a well-known problem for liquid-metal cooled fast reactors of commercial size [1]. This work investigates the performance of liquid-metal cooled fast reactors in {sup 232}Th-{sup 233}U fuel cycle and draws a comparison with the traditional {sup 238}U-{sup 239}Pu cycle. Four coolants have been considered: Na, Pb, Mg(17%at.)-Pb and Li(17%at.)-Pb; a simulation of their use in cores ranging from 700 MWth to 3600 MWth has been performed in two-dimensional diffusion theory using the European system of codes ERANOS [2,3] developed at CEA. The performance parameters such as the breeding ratio have been computed for each concept, alongside safety-related parameters: the delayed neutron fraction, the cycle reactivity swing, the Doppler constant and other thermal feedbacks. More specifically, the issue of void reactivity is studied in detail using perturbation theory. These calculations are performed at equilibrium fuel composition and are complemented by the study of the initial fuel loading at start-up which is a mixture of {sup 232}Th-{sup 239}Pu. The isotopic composition of the fissile corresponds to the plutonium available from French reactors in 2035. The conclusions of this work are that near-zero to large negative void reactivity effects can be achieved in

  3. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  4. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors; Analyse d'images tridimensionnelles ultrasonores pour l'inspection en service des reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Dancre, M

    1999-11-01

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  5. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  6. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  7. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  8. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  9. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, R.; Passerini, S.; Vilim, R. B.

    2016-04-17

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  10. On the transmutation of Am in a fast lead-cooled system

    Indian Academy of Sciences (India)

    B P Kochurov; V N Konev; A Yu Kwaretzkheli

    2007-02-01

    Characteristics of the equilibrium fuel cycle for the core or a blanket of ADS having the structure of the core of a fast lead-cooled reactor of type BREST (Russian abbreviation for `Bystryy Reaktor so Svintsovym Teplonositelem') in a mode of americium transmutation are calculated. Americium loading was taken 5% of heavy atoms. Keeping the average multiplication factor the same as in a standard equilibrium cycle, reactivity swing over 1 year's microcycle is about 1%, that demands partial fuel reloading with a periodicity of about one month. For one year of operation, 61 kg of americium is destroyed, and due to increased 238Pu content, americium is mainly converted to fission products. Thus in a system of 1 GWt (thermal), 87 kg of americium can be transmuted yearly. The estimate of the reactivity void effect has shown that it increases to 0.6% almost linearly with the void fraction increasing up to 25% and reaches its maximum of 0.7% at a void fraction of about 50%. Application of similar strategy for ADS with a sub-criticality level ≈ 0.96–0.98 can essentially relax safety problems related to positive void effects.

  11. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, Takashi, E-mail: wakai.takashi@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan); Machida, Hideo; Yoshida, Shinji [TEPCO Systems Corporation, 2-37-28 Eitai, Koto-ku, Tokyo 135 0034 (Japan); Xu, Yang [Mitsubishi FBR Systems, Inc., 2-34-17 Jingumae, Shibuya-ku, Tokyo 150 0001 (Japan); Tsukimori, Kazuyuki [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan)

    2014-04-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J{sub IC}, and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins.

  12. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  13. Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Bongsuk; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement.

  14. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

  15. Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

    Full Text Available Anticipated developments in the consumer energy market have led developers of nuclear energy concepts to consider how innovations in energy technology can be adapted to meet consumer needs. Properties of molten lead or lead-bismuth alloy coolants in lead-cooled fast reactor (LFR systems offer potential advantages for reactors with passive safety characteristics, modular deployment, and fuel cycle flexibility. In addition to realizing those engineering objectives, the feasibility of such systems will rest on development or selection of fuels and materials suitable for use with corrosive lead or lead-bismuth. Three proposed LFR systems, with varying levels of concept maturity, are described to illustrate their associated fuels and materials challenges. Nitride fuels are generally favored for LFR use over metal or oxide fuels due to their compatibility with molten lead and lead-bismuth, in addition to their high atomic density and thermal conductivity. Ferritic/martensitic stainless steels, perhaps with silicon and/or oxide-dispersion additions for enhanced coolant compatibility and improved high-temperature strength, might prove sufficient for low-to-moderate-temperature LFRs, but it appears that ceramics or refractory metal alloys will be necessary for higher-temperature LFR systems intended for production of hydrogen energy carriers.

  16. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  17. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear and Reactor Physics Laboratory, FMIPA, ITB, Physics Buildings, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  18. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  19. 用于池式快堆系统分析的钠池三维模型开发%Development of Three-Dimensional Sodium Pool Model for System Analysis of Pool-Type Liquid Metal Fast Breeder Reactor

    Institute of Scientific and Technical Information of China (English)

    隋丹婷; 陆道纲; 张盼

    2012-01-01

    由于池式快堆钠池内的热工水力学特性对反应堆的安全运行有重要影响,本文采用基于交错网格的SIMPLE算法开发直角坐标系和柱坐标系下钠池三维计算软件.应用CFX软件进行验证之后,完成了三维流场分析程序与系统分析软件SAC-CFR的耦合,并用耦合后的程序分析日本文殊快堆45%功率稳态运行工况上腔室内的流场分布,初步验证了堆芯上腔三维化的SAC-CFR用于系统分析的有效性,为进一步开发事故模型、非能动余热排出系统模型做准备.%As the thermal-hydraulic characteristic in sodium pool is crucial for safety operation of liquid metal fast breeder reactor (LMFBR), a three-dimensional sodium pool thermal-hydraulic analysis code was developed based on SIMPLE algorithm on stagger grid under Cartesian coordinates and cylindrical coordinates. After the validation with CFX, coupling between the analysis code and SAC-CFR was completed) and then the coupled code was applied to the flow field analysis in upper plenum of Monju Plant at 45% thermal power steady-state operation condition, which preliminary shows the effectiveness of the system analysis with coupled code and makes preparations for further development of accident analysis model and passive residual heat removal system.

  20. Numerical investigations of transient heat transfer characteristics and vitrification tendencies in ultra-fast cell cooling processes.

    Science.gov (United States)

    Jiao, Anjun; Han, Xu; Critser, John K; Ma, Hongbin

    2006-06-01

    During freezing, cells are often damaged directly or indirectly by ice formation. Vitrification is an alternative approach to cryopreservation that avoids ice formation. The common method to achieve vitrification is to use relatively high concentrations of cryoprotectant agents (CPA) in combination with a relatively slow cooling rate. However, high concentrations of CPAs have potentially damaging toxic and/or osmotic effects on cells. Therefore, establishing methods to achieve vitrification with lower concentrations of CPAs through ultra-fast cooling rates would be advantageous in these aspects. These ultra-fast cooling rates can be realized by a cooling system with an ultra-high heat transfer coefficient (h) between the sample and coolant. The oscillating motion heat pipe (OHP), a novel cooling device utilizing the pressure change to excite the oscillation motion of the liquid plugs and vapor bubbles, can significantly increase h and may fulfill this aim. The current investigation was designed to numerically study the effects of different values of h on the transient heat transfer characteristics and vitrification tendencies of the cell suspension during the cooling processes in an ultra-thin straw (100 microm in diameter). The transient temperature distribution, the cooling rate and the volume ratio (x) of the ice quantity to the maximum crystallizable ice of the suspension were calculated. From these numerical results, it is concluded that the ultra-high h (>10(4) W/m2 K) obtained by OHPs could facilitate vitrification by efficiently decreasing x as well as the time to pass through the dangerous temperature region where the maximum ice formation happens. For comparison, OHPs can decrease both of the parameters to less than 20% of those from the widely used open pulled straw methods. Therefore, the OHP method will be a promising approach to improving vitrification tendencies of CPA solutions and could also decrease the required concentration of CPAs for

  1. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  2. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  3. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube.

  4. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  5. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  6. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  7. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  8. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  9. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  10. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  11. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  12. Effects of mixed-method cooling on recovery of medium-fast bowling performance in hot conditions on consecutive days.

    Science.gov (United States)

    Minett, Geoffrey M; Duffield, Rob; Kellett, Aaron; Portus, Marc

    2012-01-01

    This investigation examined physiological and performance effects of cooling on recovery of medium-fast bowlers in the heat. Eight, medium-fast bowlers completed two randomised trials, involving two sessions completed on consecutive days (Session 1: 10-overs and Session 2: 4-overs) in 31 ± 3°C and 55 ± 17% relative humidity. Recovery interventions were administered for 20 min (mixed-method cooling vs. control) after Session 1. Measures included bowling performance (ball speed, accuracy, run-up speeds), physical demands (global positioning system, counter-movement jump), physiological (heart rate, core temperature, skin temperature, sweat loss), biochemical (creatine kinase, C-reactive protein) and perceptual variables (perceived exertion, thermal sensation, muscle soreness). Mean ball speed was higher after cooling in Session 2 (118.9 ± 8.1 vs. 115.5 ± 8.6 km · h⁻¹; P = 0.001; d = 0.67), reducing declines in ball speed between sessions (0.24 vs. -3.18 km · h⁻¹; P = 0.03; d = 1.80). Large effects indicated higher accuracy in Session 2 after cooling (46.0 ± 11.2 vs. 39.4 ± 8.6 arbitrary units [AU]; P = 0.13; d = 0.93) without affecting total run-up speed (19.0 ± 3.1 vs. 19.0 ± 2.5 km · h⁻¹; P = 0.97; d = 0.01). Cooling reduced core temperature, skin temperature and thermal sensation throughout the intervention (P = 0.001-0.05; d = 1.31-5.78) and attenuated creatine kinase (P = 0.04; d = 0.56) and muscle soreness at 24-h (P = 0.03; d = 2.05). Accordingly, mixed-method cooling can reduce thermal strain after a 10-over spell and improve markers of muscular damage and discomfort alongside maintained medium-fast bowling performance on consecutive days in hot conditions.

  13. Stability analysis of a natural circulation lead-cooled fast reactor

    Science.gov (United States)

    Lu, Qiyue

    This dissertation is aimed at nuclear-coupled thermal hydraulics stability analysis of a natural circulation lead cooled fast reactor design. The stability concerns arise from the fact that natural circulation operation makes the system susceptible to flow instabilities similar to those observed in boiling water reactors. In order to capture the regional effects, modal expansion method which incorporates higher azimuthal modes is used to model the neutronics part of the system. A reduced order model is used in this work for the thermal-hydraulics. Consistent with the number of heat exchangers (HXs), the reactor core is divided into four equal quadrants. Each quadrant has its corresponding external segments such as riser, plenum, pipes and HX forming an equivalent 1-D closed loop. The local pressure loss along the loop is represented by a lumped friction factor. The heat transfer process in the HX is represented by a model for the coolant temperature at the core inlet that depends on the coolant temperature at the core outlet and the coolant velocity. Additionally, time lag effects are incorporated into this HX model due to the finite coolant speed. A conventional model is used for the fuel pin heat conduction to couple the neutronics and thermal-hydraulics. The feedback mechanisms include Doppler, axial/radial thermal expansion and coolant density effects. These effects are represented by a linear variation of the macroscopic cross sections with the fuel temperature. The weighted residual method is used to convert the governing PDEs to ODEs. Retaining the first and second modes, leads to six ODEs for neutronics, and five ODEs for the thermal-hydraulics in each quadrant. Three models are developed. These are: 1) natural circulation model with a closed coolant flow path but without coupled neutronics, 2) forced circulation model with constant external pressure drop across the heated channels but without coupled neutronics, 3) coupled system including neutronics with

  14. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  15. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  16. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    OpenAIRE

    Gottfridsson, Filip

    2010-01-01

    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  17. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  18. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  19. Fast heating and cooling in nanoimprint using a spring-loaded adapter in a preheated press

    DEFF Research Database (Denmark)

    Schift, Helmut; Bellini, Sandro; Gobrecht, Jens

    2007-01-01

    By using a spring-loaded adapter, instant heating and cooling of wafer-type substrates was implemented in standard hot embossing equipment. This was possible by using the well-known concept of a clamped stack of stamp and substrate, pre-assembled in an alignment fixture. A number of thermoplastic...

  20. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    Science.gov (United States)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  1. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

  2. Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN Fuel Long Life without Refueling

    Directory of Open Access Journals (Sweden)

    Syarifah Ratna Dewi

    2016-01-01

    Full Text Available Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

  3. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  4. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    OpenAIRE

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    International audience; The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into acc...

  5. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  6. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    Science.gov (United States)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  7. Investigations on the heat transport capability of a cryogenic oscillating heat pipe and its application in achieving ultra-fast cooling rates for cell vitrification cryopreservation☆

    OpenAIRE

    Han, Xu; Ma, Hongbin; Jiao, Anjun; Critser, John K.

    2008-01-01

    Theoretically, direct vitrification of cell suspensions with relatively low concentrations (~1 M) of permeating cryoprotective agents (CPA) is suitable for cryopreservation of almost all cell types and can be accomplished by ultra-fast cooling rates that are on the order of 106–7 K/min. However, the methods and devices currently available for cell cryopreservation cannot achieve such high cooling rates. In this study, we constructed a novel cryogenic oscillating heat pipe (COHP) using liquid ...

  8. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  9. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  10. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  11. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  12. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Science.gov (United States)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  13. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  14. Preliminary Study of Lead-Oxide Cooled Fast Reactor with Natural Uranium as an Input Fuel with Reactor Shuffling Strategy

    Science.gov (United States)

    Mahmudah, Rida SN; Su’ud, Zaki

    2017-01-01

    A preliminary study of lead-oxide cooled fast reactor with natural uranium as an input fuel using reactor shuffling strategy has been conducted. In this study, reactor core is divided into four zone with the same volume, each zone use different uranium enrichment. The enrichment number is estimated so that in the end of reactor’s operation, we only need to add natural uranium as the fresh input fuel. This study used UN-PuN as the fuel and lead oxide as the coolant. Several parameter studies have been conducted to determine the most suitable input condition. It is confirmed in this study that with fuel : cladding : coolant ratio of 53 : 10 : 37, and uranium enrichment in the first to the fourth zone of 0%, 6.25%, 7.5% and 8%, respectively, the reactor can operate as long as 20 years of operation with terminal k-eff of 1.0004.

  15. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  16. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  17. Microstructure and Property of Mn-Nb-B Low Carbon Bainite High Strength Steel Under Ultra-fast Cooling

    Directory of Open Access Journals (Sweden)

    WANG Bing-xing

    2016-07-01

    Full Text Available Using the Mn-Nb-B low carbon bainite high strength steel with the reducing production technology as the research target, the deformation behavior and phase transformation behavior were studied by the thermal simulation testing machine. Combining with the characteristics of the medium and heavy plate production line, the controlled rolling and controlled cooling technology based on ultra-fast cooling were designed to produce low cost high strength construction machinery steel with superior comprehensive mechanical properties. The strengthening mechanisms such as grain refinement strengthening, precipitation strengthening are effective to produce the Mn-Nb-B low carbon bainite high strength steel. The yield strength and tensile strength of the product reach to 678MPa and 756 MPa respectively, the elongation A50 is 33% and the impact energy at -20℃ is 261J. The microstructure of the steel is composed of granular bainite, acicular ferrite and lath bainite. A large number of fine, point, granular M/A constituents and dislocation structures dispersively distributed inside the matrix, and also tiny and dispersed (Nb,Ti (C,N precipitates are observed by transmission electron microscopy.

  18. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  19. Resfriamento rápido de fornos de carbonização Fast cooling of mud ovens of carbonization

    Directory of Open Access Journals (Sweden)

    Delly Oliveira Filho

    2010-12-01

    of steam, with the goal of reducing the cooling time of the charcoal. It was built an oven type container in a test scale with the following dimensions 3.0 x a 2.0 x 1.6 m, as well as a system burner for exhaustion gases from the wood carbonization, instrumented thermocouples type K and a data acquisition system at the Laboratories of the Energy Area and Pre Processing of Agricultural Products of the Agricultural Engineering Department of Federal University of Viçosa, Viçosa, Minas Gerais, Brazil. It was built an oven container type on a laboratory scale of dimensions equal to 3.0 x 2.0 x 1.6 m and a firing system of gas exhaustion from the carbonization of wood, instrumented with temperature sensors type k, connected to a data system acquisition. It was also used a boiler to generate steam for the fast cooling of the oven. The cooling time of the coal mass was reduced to about two days by the injection of water vapor when compared to the natural cooling process.

  20. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  1. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  2. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  3. Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

    Energy Technology Data Exchange (ETDEWEB)

    Riber Marklund, A. [CEA, Cadarache, DEN/DTN/STCP/LIET, Batiment 202, 13108 St Paul-lez-Durance, (France); Kishore, S. [Fast Reactor Technology Group of IGCAR, (India); Prakash, V. [Vibrations Diagnostics Division, Fast Reactor Technology Group of IGCAR, (India); Rajan, K.K. [Fast Reactor Technology Group and Engineering Services Group of IGCAR, (India)

    2015-07-01

    Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)

  4. Safety properties of sodium-cooled fast reactors%钠冷快堆及其安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤; 杨红义

    2016-01-01

    钠冷快堆是第四代核能系统国际论坛(GIF)公布的6种第四代先进反应堆中研发进展最快、最接近满足商业核电厂需要的堆型。钠冷快堆因其在固有安全性以及可增殖核燃料、嬗变长寿命放射性废物等方面的优势,得到了世界各国的重视。文章以中国第一座钠冷快堆——中国实验快堆(China Experimental Fast Reactor,CEFR)为例,介绍了钠冷快堆在设计及运行方面的安全特性。%The sodium-cooled fast reactor is the fastest prototype and the closest to com-mercialization for nuclear power plants amongst the six types of fourth generation reactors, as an-nounced at the Generation IV International Forum. Many countries are paying more and more at-tention to the research and development of these reactors, due to the inherent safety features, effi-cient utilization of uranium with the breeding of the plutonium, and transmutation of long-lived ac-tinides. The design and operational safety characteristics of the China Experimental Fast Reactor are reviewed in this paper.

  5. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  6. Efeitos das condições ambientais no período entre a postura e o armazenamento de ovos de matrizes pesadas sobre o rendimento de incubação Effect on time between oviposition and the beginning of cool storage on hatchability of broiler breeder eggs

    Directory of Open Access Journals (Sweden)

    M.A. Fiuza

    2006-06-01

    Full Text Available Avaliou-se o efeito das condições ambientais no período entre a postura e o armazenamento de ovos sobre o rendimento de incubação de 2.742 ovos de matrizes pesadas da linhagem Ross com 31 semanas de idade. Os tratamentos foram definidos pelo período de permanência dos ovos sob temperatura ambiente do galpão antes do armazenamento. Para cada tratamento foram utilizados 860 ovos, distribuídos em 10 repetições de 86 ovos cada. No tratamento A, os ovos foram enviados para a sala fria imediatamente após a coleta; no B, foram enviados para a sala fria após cinco horas de permanência no galpão; e, no C, permaneceram no galpão por 10 horas, antes de serem enviados para a sala fria. As condições ambientais da sala fria foram: temperatura 18,2 a 21,0 ºC e umidade relativa do ar de 72,8 a 76,8%. O período de armazenamento dos ovos foi de quatro dias. A mortalidade embrionária no tratamento A foi maior que no tratamento B (PThe effect on hatchability of time post-oviposition until initiation of cool storage was examined using 2,742 eggs from 31-week-old Ross hens. Treatments were defined by the length of time that eggs remained at breeder house temperature, prior to cool storage. In treatments A, B and C, respectively, eggs were taken to the cooling room immediately after collection, after 5 hours in the aviary or after 10 hours in the aviary. From each treatment, 860 eggs were distributed into 10 replicates of 86 eggs each and stored for 4 days at 18.2 - 21.0ºC and 75% humidity. After cool storage, they were incubated. Embryo mortality in treatment A was higher than in B (P<0.05. The hatchability of eggs of treatment B was higher (P<0.05 than that of treatment A. Average chick weight at hatching was higher for treatment A than for treatment C (P<0.05. Eggs that were cooled beginning five hours after oviposition produced the highest incubation yield.

  7. Literature review on metallic fuel source term for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Nam Duk; Bae, Moo Hoon; Shin, An Dong; Huh, Chang Wook [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2012-10-15

    Source term is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment, following a severe accident where a significant portion of the reactor core has melted. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. Apart from assessing the radiological consequences for siting, it is also important for designing filtering systems and even reactor components. Overly conservative source term for light water reactor, TID 14844 demands for very fast closure of main steam isolation valves, rapid startup of emergency diesels, and safety systems designed to mitigate gaseous iodine. In spite of this importance, most of the knowledge we have for SFR source term comes from the research performed before 1980s. Moreover, majority of the work on metallic fuels was done during the late 1950's through the 1960's. This paper reviews and summarizes the main characteristics of SFR source terms based on the available literatures.

  8. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  9. Highly c-axis-oriented monocrystalline Pb(Zr, Ti)O₃ thin films on si wafer prepared by fast cooling immediately after sputter deposition.

    Science.gov (United States)

    Yoshida, Shinya; Hanzawa, Hiroaki; Wasa, Kiyotaka; Esashi, Masayoshi; Tanaka, Shuji

    2014-09-01

    We successfully developed sputter deposition technology to obtain a highly c-axis-oriented monocrystalline Pb(Zr, Ti)O3 (PZT) thin film on a Si wafer by fast cooling (~-180°C/min) of the substrate after deposition. The c-axis orientation ratio of a fast-cooled film was about 90%, whereas that of a slow-cooled (~-40°C/min) film was only 10%. The c-axis-oriented monocrystalline Pb(Zr0.5, Ti0.5)O3 films showed reasonably large piezoelectric coefficients, e(31,f) = ~-11 C/m(2), with remarkably small dielectric constants, ϵ(r) = ~220. As a result, an excellent figure of merit (FOM) was obtained for piezoelectric microelectromechanical systems (MEMS) such as a piezoelectric gyroscope. This c-axis orientation technology on Si will extend industrial applications of PZT-based thin films and contribute further to the development of piezoelectric MEMS.

  10. Modeling of natural circulation for the inherent safety analysis of sodium cooled fast reactors

    Directory of Open Access Journals (Sweden)

    A.S. Bochkarev

    2016-12-01

    Full Text Available The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor. The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed. It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat sink and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor. The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system's intermediate circuit and by the heat transfer through the reactor vessel. The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety. The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.

  11. Optimization and application of cooled avalanche photodiodes for spectroscopic fluctuation measurements with ultra-fast charge exchange recombination spectroscopy

    Science.gov (United States)

    Truong, D. D.; Fonck, R. J.; McKee, G. R.

    2016-11-01

    The Ultra-Fast Charge Exchange Recombination Spectroscopy (UF-CHERS) diagnostic is a highly specialized spectroscopic instrument with 2 spatial channels consisting of 8 spectral channels each and a resolution of ˜0.25 nm deployed at DIII-D to measure turbulent ion temperature fluctuations. Charge exchange emissions are obtained between 528 and 530 nm with 1 μs time resolution to study plasma instabilities. A primary challenge of extracting fluctuation measurements from raw UF-CHERS signals is photon and electronic noise. In order to reduce dark current, the Avalanche Photodiode (APD) detectors are thermo-electrically cooled. State-of-the-art components are used for the signal amplifiers and conditioners to minimize electronic noise. Due to the low incident photon power (≤1 nW), APDs with a gain of up to 300 are used to optimize the signal to noise ratio. Maximizing the APDs' gain while minimizing the excess noise factor (ENF) is essential since the total noise of the diagnostic sets a floor for the minimum level of detectable broadband fluctuations. The APDs' gain should be high enough that photon noise dominates electronic noise, but not excessive so that the ENF overwhelms plasma fluctuations. A new generation of cooled APDs and optimized preamplifiers exhibits significantly enhanced signal-to-noise compared to a previous generation. Experiments at DIII-D have allowed for characterization and optimization of the ENF vs. gain. A gain of ˜100 at 1700 V is found to be near optimal for most plasma conditions. Ion temperature and toroidal velocity fluctuations due to the edge harmonic oscillation in quiescent H-mode plasmas are presented to demonstrate UF-CHERS' capabilities.

  12. Mitigation of corrosion and mass transfer in sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Latge, C. [CEA Cadarache, Dir. de l' Energie Nucleaire, 13 - Saint-Paul-lez-Durance (France); Feron, D. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif-sur-Yvette (France)

    2009-07-01

    Full text of publication follows: Several coolants can be used for the development of the Fast Reactors, as sodium, gas, lead or lead-bismuth eutectic, and have been selected in the Generation IV forum. The high density energy requires a coolant with a very good thermal conductivity. Liquid sodium is such a medium which is liquid between 97.8 up to 880 C at dynamic pressure below 4 bars, and with compatible neutron-physical properties. Its viscosity is comparable to that of water and its compatibility with metallic materials is fairly satisfactory. It is however necessary to keep the conditions of operation within a range such that corrosion is limited. Several materials are suitable for use in liquid sodium reactors, among ferritic and austenitic steels and high temperature alloys with up to 32% nickel contents. The designer has however to consider the mass transfer between materials of different compositions. The exchange and transfer of non-metallic elements such as carbon or nitrogen has to be taken into account. The corrosion mechanisms of austenitic steels have been extensively studied and described in the literature: surface cleaning, austenitic dissolution, formation of a ferrite layer, steady state equilibrium and several models have been proposed: main parameters include oxygen content, sodium velocity and steel temperature. Operating experience has shown that, if there are no cladding failures, the main source of radioactivity in the primary circuit is the activated corrosion products, like {sup 54}Mn, {sup 51}Cr,..., induced by the activation of core materials which are dissolved into the sodium and mainly deposited in the coldest parts of the reactor i.e. the Intermediate Heat Exchanger (IHX) and pumps. Radio-cobalt such as {sup 60}Co are also produced and a low fraction is deposited in primary components. The corrosion rates estimated and the contamination induced by activated corrosion products observed in SFR like Phenix, JOYO, BN600, PFR, EBR2 have

  13. Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

    2016-04-15

    Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

  14. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  15. Stability analysis of the Korean prototype generation-IV sodium-cooled fast reactor using linear frequency domain approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ji; Ha, Pham Nhu Viet; Lim, Jae Yong; Hahn, Do Hee; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of). Fast Reactor Development Div.

    2016-03-15

    The Korea Atomic Energy Research Institute (KAERI) has been developing the 150 MWe Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR). The design concept is highly based on passive safety mechanisms, minimizing the need for engineered safety systems. Presently, it is of primary importance to assure the reactor dynamics and stability against small reactivity disturbances under power operating conditions. KAERI has therefore developed the NuSTAB code for stability analysis of the PGSFR. In NuSTAB, the neutron-kinetic and thermal-hydraulic coupling equations are linearized to form the characteristic equation, which is solved as a generalized eigenvalue problem for determining the decay ratio, an indicator of the system stability. In this paper, the stability of the PGSFR was analyzed by applying the point kinetic and spatial kinetic options in the NuSTAB code. System responses to temperature feedbacks including the Doppler effect, thermal expansion, coolant density change, and overall feedback were studied. The results indicate that the initial U and final TRU cores of the PGSFR are both inherently stable thanks to the temperature feedbacks.

  16. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; 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 Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  17. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable.

  18. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  19. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sham, Sam [ORNL; Tan, Lizhen [ORNL; Yamamoto, Yukinori [ORNL

    2013-01-01

    Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for the advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.

  20. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  1. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  2. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  3. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  4. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  5. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  6. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  7. Breeder nutrition and offspring performance

    Directory of Open Access Journals (Sweden)

    F Calini

    2007-06-01

    Full Text Available Vertical integration in poultry industry strongly emphasizes the importance of cost control at all levels. In the usual broiler production operations, the costs involved with the production of the hatching egg or the day old chick are negligible if seen in the perspective of the cost per kg of live bird. From a research point of view, anyway, the greatest attention is usually given to the performance of broiler breeders, and most of the research in the field is focused on the improvement of their relative performance, mainly in terms of saleable chicks produced per hen, while less attention has been given to the quality of the chick and to the improvement of its growth performances, even if these last parameters have an effective impact on the overall economics of the poultry growing business. Most of the data available is quite dated, as can be seen from some recent reviews, and in general little attention is given to the impact of parental nutrition on the subsequent broiler performance. It is in fact more usual to find data about dam nutrition influence on egg fertility and hatchability than on subsequent progeny performance. The objectives of this review were to assess, on the basis of published reports, the effects of selected nutrients and anti-nutrients normally prevailing in commercial broiler breeder feeds - vitamins, micro-minerals, mycotoxins, - trying to pinpoint which could be the positive and the negative effects of both on the subsequent broiler performance, with a particular attention to the impact on immune function and carcass yield.

  8. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  9. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  10. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  11. Constrains on Crustal Accretion Obtained from Cooling Rate Calculations with a Thermo-Mechanical Model of Fast-Spreading Mid-Ocean Ridges

    Science.gov (United States)

    Garrido, C. J.; Machetel, P.

    2012-12-01

    We have used a thermo-mechanical model designed to find steady-state solutions of motion and temperature with variable viscosity, heat diffusion, heat advection, hydrothermal cooling and latent heat release. Cases analogous to the "gabbro glacier" (G accretion structure), "sheeted sills" (S structure) and "mixed shallow and MTZ lenses" (M structure) were computed with and without sheeted dyke level modeling. The results show that thermal patterns near the ridge mainly depend on hydrothermal cooling. Several hydrothermal cooling cracking temperature have been used in order to illustrate the present scientific debate on the penetration depth and efficiency of hydrothermal flows. Second, higher cooling rates are obtained for the G structures. Third, whereas the subsolidus cooling rates, SCR, decrease monotonically with depth, the igneous cooling rates, ICR, display local minima at the merging levels of the upper and lower lenses. It appears that ICR reveal the near-ridge thermal and mechanical structures, whereas the lower value of the initial-to-closure temperature ranges used for SCR cause shifts farther from the ridge that reduces the ability of SCR to discriminate the ridge thermo-mechanical configuration. It also indicates that the common assumption that ICR and SCR should be similar is probably over-simplified. Finally, the cooling rates obtained bears the clear signature of the three intrusion hypothesis. The results show that numerical modeling of the lower crust's thermo-mechanical properties may provide new insights to discriminate among hypotheses related to G, M and S structures for fast-spreading ridges.; Thermal history obtained for the Gabro Glacier (top panels), Mixed shallow and MTZ zone (middle panels) and Sheeted Sills hypothesis (bottom panels)for the magma intrusion at ridge. Columns corresponds to various hydrothermal cooling and viscosity hypothesis.

  12. 堆内辐照在线产氚氦冷陶瓷氚增殖剂包层球床组件热工流体力学计算分析%Thermal-hydraulic Calculation and Analysis on Helium Cooled Ceramic Breeder Pebble Bed Assembly for In-pile Irradiation & In-situ Tritium Extraction

    Institute of Scientific and Technical Information of China (English)

    郭春秋; 解家春; 刘兴民

    2013-01-01

    在研究堆内进行辐照在线产氚试验是ITER计划专项国内配套研究项目之一.本工作主要针对研究堆内辐照氦冷陶瓷氚增殖剂包层(简称陶瓷)球床组件的试验技术要求,评估辐照陶瓷球床组件设计方案的可行性.通过对不同的陶瓷球床组件结构参数和组件在堆内的不同辐照位置,进行热工流体力学设计计算,得到满足要求的入堆辐照陶瓷球床组件设计方案.%In-pile irradiation & in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program.According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor,the feasibility of the design for the in-pile irradiation & in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated.By conducting thermal-hydraulic design calculation with different in-pile irradiation channels,locations and structure parameters for ceramic pebble bed assembly,a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained.

  13. Vitrification by Ultra-fast Cooling at a Low Concentration of Cryoprotectants in a Quartz Microcapillary: A Study Using Murine Embryonic Stem Cells

    OpenAIRE

    He, Xiaoming; Park, Eric Y H; Fowler, Alex; Martin L. Yarmush; Toner, Mehmet

    2008-01-01

    Conventional cryopreservation protocols for slow-freezing or vitrification involve cell injury due to ice formation/cell dehydration or toxicity of high cryoprotectant (CPA) concentrations, respectively. In this study, we developed a novel cryopreservation technique to achieve ultra-fast cooling rates using a quartz microcapillary (QMC). The QMC enabled vitrification of murine embryonic stem (ES) cells using an intracellular cryoprotectant concentration in the range used for slowing freezing ...

  14. The European JASMIN Project for the Development of a New Safety Simulation Code, ASTEC-Na, for Na-cooled Fast Neutron Reactors

    OpenAIRE

    GIRAULT N.; VAN DORSSELAERE J.p.; Jacq, F.; BRILLANT G.; KISSANE Martin; BANDINI, G; Buck,M.; CHAMPIGNY J.; Hering, W; Perez-Martin, S.; Herranz, L; RAISON Philippe; Reinke, N; TUCEK Kamil; VERWAERDE D.

    2012-01-01

    The 4-year JASMIN collaborative project, involving 9 organizations, was launched by IRSN end of 2011 within the 7th European R&D Framework Programme on the enhancement of Na-cooled Fast Neutron Reactors (SFR) safety for a higher resistance to severe accidents. The project aims at developing a new European simulation code, ASTEC-Na, with a modern architecture, sufficiently flexible to account for innovative reactor designs and eventually new types of fuel and claddings and accounting for resul...

  15. Investigations on the heat transport capability of a cryogenic oscillating heat pipe and its application in achieving ultra-fast cooling rates for cell vitrification cryopreservation.

    Science.gov (United States)

    Han, Xu; Ma, Hongbin; Jiao, Anjun; Critser, John K

    2008-06-01

    Theoretically, direct vitrification of cell suspensions with relatively low concentrations ( approximately 1 M) of permeating cryoprotective agents (CPA) is suitable for cryopreservation of almost all cell types and can be accomplished by ultra-fast cooling rates that are on the order of 10(6-7) K/min. However, the methods and devices currently available for cell cryopreservation cannot achieve such high cooling rates. In this study, we constructed a novel cryogenic oscillating heat pipe (COHP) using liquid nitrogen as its working fluid and investigated its heat transport capability to assess its application for achieving ultra-fast cooling rates for cell cryopreservation. The experimental results showed that the apparent heat transfer coefficient of the COHP can reach 2 x 10(5) W/m(2).K, which is two orders of the magnitude higher than traditional heat pipes. Theoretical analyzes showed that the average local heat transfer coefficient in the thin film evaporation region of the COHP can reach 1.2 x 10(6) W/m(2).K, which is approximately 10(3) times higher than that achievable with standard pool-boiling approaches. Based on these results, a novel device design applying the COHP and microfabrication techniques is proposed and its efficiency for cell vitrification is demonstrated through numerical simulation. The estimated average cooling rates achieved through this approach is 10(6-7)K/min, which is much faster than the currently available methods and sufficient for achieving vitrification with relatively low concentrations of CPA.

  16. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  17. FAST

    DEFF Research Database (Denmark)

    Zuidmeer-Jongejan, Laurian; Fernandez-Rivas, Montserrat; Poulsen, Lars K.

    2012-01-01

    ABSTRACT: The FAST project (Food Allergy Specific Immunotherapy) aims at the development of safe and effective treatment of food allergies, targeting prevalent, persistent and severe allergy to fish and peach. Classical allergen-specific immunotherapy (SIT), using subcutaneous injections with aqu......ABSTRACT: The FAST project (Food Allergy Specific Immunotherapy) aims at the development of safe and effective treatment of food allergies, targeting prevalent, persistent and severe allergy to fish and peach. Classical allergen-specific immunotherapy (SIT), using subcutaneous injections...... with aqueous food extracts may be effective but has proven to be accompanied by too many anaphylactic side-effects. FAST aims to develop a safe alternative by replacing food extracts with hypoallergenic recombinant major allergens as the active ingredients of SIT. Both severe fish and peach allergy are caused...... in depth serological and cellular immune analyses will be performed, allowing identification of novel biomarkers for monitoring treatment efficacy. FAST aims at improving the quality of life of food allergic patients by providing a safe and effective treatment that will significantly lower their threshold...

  18. Irradiation behavior of metallic fast reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  20. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  1. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  2. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  3. Concept of a Staged FEL Enabled by Fast Synchrotron Radiation Cooling of Laser-Plasma Accelerated Beam by Solenoidal Magnetic Fields in Plasma Bubble

    CERN Document Server

    Seryi, Andrei; Andreev, Alexander; Konoplev, Ivan

    2016-01-01

    A novel method for generating GigaGauss solenoidal field in laser-plasma bubble, using screw-shaped laser pulses, has been recently presented in arXiv:1604.01259 [physics.plasm-ph]. Such magnetic fields enable fast synchrotron radiation cooling of the beam emittance of laser-plasma accelerated leptons. This recent finding opens a novel approach for design of laser-plasma FELs or colliders, where the acceleration stages are interleaved with laser-plasma emittance cooling stages. In this concept paper we present an outline of how a staged plasma-acceleration FEL could look like and discuss further studies needed to investigate the feasibility of the concept in detail.

  4. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  5. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Science.gov (United States)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Aziz, Ferhat; Permana, Sidik; Sekimoto, Hiroshi

    2014-02-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  6. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs.

  7. Design study of an upgraded charge breeder for ISOLDE

    CERN Document Server

    Shornikov, A; Wenander, F; Pikin, A

    2013-01-01

    In this work we present our progress in the design study of a new Electron Beam Ion Source (EBIS) to be installed as a charge breeder for reacceleration of rare ions at ISOLDE. The work is triggered by the HIE-ISOLDE upgrade {[}1] and the planned TSR@ISOLDE project {[}2]. To fulfill the requests of the user community the new EBIS should reach an electron beam density of 10(4) A/cm(2) at electron energies up to 150 key and, provide UHV environment and ion cooling in the breeding region to ensure confinement of the ions long enough to reach the requested charge states. We report on the established design parameters and first prototyping steps towards production and testing of suitable equipment. (C) 2013 Elsevier B.V. All rights reserved.

  8. Evaluation of nutrient specifications for broiler breeders.

    Science.gov (United States)

    Wilson, H R; Harms, R H

    1984-07-01

    Two experiments were conducted to determine if previously suggested nutrient requirements of broiler breeders (23 g protein, 850 mg sulfur amino acids, 4.5 g calcium, and 750 mg phosphorus/bird/day) are in excess and could be reduced during the laying period. In Experiment 1, Cobb color-sex broiler breeders were fed daily nutrient allowances that were 100.0, 96.3, 92.5, 89.4, and 86.6% of the suggested requirements. In Experiment 2, Cobb feather-sex breeders were fed daily allowances that were 92.5, 89.4, 86.6, 83.4, and 80.9% of the suggested requirements. Birds on all diets were fed the same energy level; however, energy varied with season to maintain body weight. Egg production, fertility, hatchability, egg weight, and shell quality were not significantly affected by the reductions in nutrient intake in either experiment. The diet with the lowest nutrient level (80.9% of the suggested requirement) was adequate, indicating a considerable margin of safety for the stated requirements. Body weight was quite variable but tended to decrease with nutrient restriction. Weights of broilers hatched from treated breeders were not significantly affected at 49 days of age by the breeder dietary treatments. These results indicate that broiler breeder diets formulated to meet presently suggested requirements have a large margin of safety and a reduction of specifications by approximately 10% is suggested. The revised daily intakes recommended are: 20.6 g protein, 754 mg sulfur amino acids, 400 mg methionine, 938 mg lysine, 1379 mg arginine, 256 mg tryptophan, 4.07 g calcium, 683 mg total phosphorus, and 170 mg sodium.

  9. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor; Spectrometrie gamma haute resolution et hauts taux de comptage sur primaire de reacteur de type generation 4 au sodium liquide

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.

    2010-11-10

    Sodium-cooled Fast Reactors are under development for the fourth generation of nuclear reactor. Breeders reactors could gives solutions for the need of energy and the preservation of uranium resources. An other purpose is the radioactive wastes production reduction by transmutation and the control of non-proliferation using a closed-cycle. These thesis shows safety and profit advantages that could be obtained by a new generation of gamma spectrometry system for SFR. Now, the high count rate abilities, allow us to study new methods of accurate power measurement and fast clad failure detection. Simulations have been done and an experimental test has been performed at the French Phenix SFR of the CEA Marcoule showing promising results for these new measurements. (author) [French] Les reacteurs a neutrons rapides refroidis au sodium sont en developpement en vue d'assurer une quatrieme generation de reacteurs repondant a la demande energetique, tout en assurant la preservation des ressources d'uranium par un fonctionnement en surgenerateur. L'objectif de la filiere est egalement d'ameliorer la gestion de la radiotoxicite des dechets produits par transmutation des actinides mineurs et de controler la non-proliferation par un fonctionnement en cycle ferme. Une instrumentation de surveillance et de controle de ce type de reacteur a ete etudiee dans cette these. La spectrometrie gamma de nouvelle generation permet, par les hauts taux de traitement aujourd'hui accessibles, d'envisager de nouvelles approches pour suivre avec une precision accrue la puissance neutronique et de detecter plus precocement des ruptures de gaine combustible. Des simulations numeriques ont ete realisees et une campagne d'essai a ete menee a bien sur le reacteur Phenix de Marcoule. Des perspectives prometteuses ont ete mises en exergue pour ces deux problematiques

  10. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  11. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  12. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  13. Multi-objective Optimizations of a Novel Cryo-cooled DC Gun Based Ultra Fast Electron Diffraction Beamline

    OpenAIRE

    Gulliford, C.; Bartnik, A.; Bazarov, I.

    2015-01-01

    We present the results of multi-objective genetic algorithm optimizations of a potential single shot ultra fast electron diffraction beamline utilizing a 225 kV dc gun with a novel cryocooled photocathode system and buncher cavity. Optimizations of the transverse projected emittance as a function of bunch charge are presented and discussed in terms of the scaling laws derived in the charge saturation limit. Additionally, optimization of the transverse coherence length as a function of final r...

  14. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  15. Thermal Expansion Measurements on Boron Carbide for Fast Breeder Reactor

    Institute of Scientific and Technical Information of China (English)

    1995-01-01

    1.9ThermalExpansionMeasurementsonBoronCarbideforFastBreederReactorZhangLili;HuangYingB_4Cisneutronabsorbermaterialforcontrolr...

  16. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    Science.gov (United States)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  17. Formation of very hard electron and gamma-ray spectra of flat-spectrum radio quasars in the fast-cooling regime

    Science.gov (United States)

    Yan, Dahai; Zhang, Li; Zhang, Shuang-Nan

    2016-07-01

    In the external Compton scenario, we investigate the formation of a very hard electron spectrum in the fast-cooling regime, using a time-dependent emission model. It is shown that a very hard electron distribution, N^' }_e({γ ^' })∝ {γ ^' }^{-p}, with spectral index p ˜ 1.3 is formed below the minimum energy of injection electrons when inverse Compton scattering takes place in the Klein-Nishina regime, i.e. inverse Compton scattering of relativistic electrons on broad-line region radiation in flat-spectrum radio quasars. This produces a very hard gamma-ray spectrum and can explain in reasonable fashion the very hard Fermi-Large Area Telescope (LAT) spectrum of the flat-spectrum radio quasar 3C 279 during the extreme gamma-ray flare in 2013 December.

  18. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  19. 中厚板超快速冷却多目标控制研究及应用%Research and application of multi-objective control of ultra fast cooling for plate

    Institute of Scientific and Technical Information of China (English)

    刘涛; 余伟; 李彦彬; 殷实; 何春雨

    2015-01-01

    The cooling temperature is the main control target for the traditional cooling system,while the control of the cooling rate and the cooling uniformity is ignored. The ultra fast cooling system is developed in this paper to realize the multi-objective control of the cooling route and the uniform cool-ing. That is to say,mathematical models are optimized and the corresponding control strategies are de-veloped to achieve the accurate control of the final cooling temperature and the cooling rate,and the uniform cooling in plate width,length and thickness directions is achieved through strategies such as edge masking,head and tail masking, micro acceleration of the roll table, symmetrical cooling, etc. The ultra fast cooling system is applied in a certain domestic iron and steel corporation. The results show that the indices such as the final cooling temperature,the cooling rate,the temperature uniformi-ty after cooling, the performance uniformity, etc. meet the control requirements, and the application effect is good.%传统的轧后控冷系统以温度作为主要控制目标,忽略了对冷却速率及冷却均匀性的控制,为此,作者开发了超快速冷却系统,实现了冷却路径及均匀冷却的多目标控制,即,通过优化数学模型算法及相应控制策略实现终冷温度及冷却速率精确控制,通过边部遮蔽、头尾遮蔽、辊道微加速及对称冷却等策略实现钢板横向、纵向及厚度方向的均匀冷却。超快冷系统在国内某钢铁公司应用后,终冷温度、冷却速率、冷却后温度均匀性及性能均匀性等指标均达到控制要求,应用效果良好。

  20. Design Improvement of Iso-Kinetic Flow Sampling Device at Subchannel in a Wire-Wrapped 37-pin Fuel Assembly for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Won; Kim, Hyungmo; Ko, Yung-Joo; Chang, Seok-Kyu; Choi, Hae Seob; Euh, Dong-kin; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Securing the structural integrity of a fuel assembly during reactor operation is of utmost importance in order to prevent reactor severe accident like the Fukushima nuclear power plant through a flow characteristics tests with test assembly scaled down from a prototype reactor of a sodium-cooled fast reactor (SFR). To evaluate uncertainty is very important to ensure reliability at the results of the fuel assembly. Therefore the sub-channel analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. In KAERI, two sub-channel analysis codes (SLTHEN, MATRA-LMR) are considered to utilize for the design of the prototype reactor. In this study, design improvement of iso-Kinetic flow sampling device at sub-channel in a wire-wrapped 37-pin fuel assembly for a sodium cooled fast reactor is conducted for decreasing misalignment sensitivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In KAERI, two subchannel analysis codes are considered to be utilized for the design of the prototype reactor. In this study, the X-axis probe misalignment error is 2.5%, the Y-axis probe misalignment error is 0.9% and flowmeter and DA equipment error is 0.2%. As shown in above results, the misalignment error was the highest factor in uncertainty analysis. To solve the problem, design improvement of iso-kinetic flow sampling device at subchannel in a wire-wrapped 37-pin fuel assembly is practiced for decreasing misalignment sensitivity error.

  1. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  2. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  3. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  4. FOWL CHOLERA IN A BREEDER FLOCK

    Directory of Open Access Journals (Sweden)

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-12-01

    Full Text Available During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  5. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  6. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  7. Design of alumina forming FeCrAl steels for lead or lead-bismuth cooled fast reactors

    Science.gov (United States)

    Lim, Jun; Hwang, Il Soon; Kim, Ji Hyun

    2013-10-01

    Iron-chromium-aluminum alloys containing 15-20 wt.% Cr and 4-6 wt.% Al have shown excellent corrosion resistance in the temperature range up to 600 °C or higher in liquid lead and lead-bismuth eutectic environments by the formation of protective Al2O3 layers. However, the higher Cr and Al concentrations in ferritic alloys could be problematic because of severe embrittlement in the manufacturing process as well as in service, caused by the formation of brittle phases. For this reason, efforts worldwide have so far mainly focused on the development of aluminizing surface treatments. However, aluminizing surface treatments have major disadvantages of cost, processing difficulties and reliability issues. In this study, a new FeCrAl alloy is proposed for structural materials in lead and lead-bismuth cooled nuclear applications. The alloy design relied on corrosion experiments in high temperature lead and lead-bismuth eutectic environments and computational thermodynamic calculations using the commercial software, JMatPro. The design of new alloys has focused on the optimization of Cr and Al levels for the formation of an external Al2O3 layer which can provide excellent oxidation and corrosion resistance in liquid lead alloys in the temperature range 300-600 °C while still retaining workable mechanical properties.

  8. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  9. What determines hatchling weight: breeder age or incubated egg weight?

    Directory of Open Access Journals (Sweden)

    AB Traldi

    2011-12-01

    Full Text Available Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval determined for each breeder age were weighed, broken, and their components were separated and weighed. At hatch, hatchlings were sexed and weighed, determining the average initial weight of the progeny of each breeder age. Data were analyzed using the Analyst program of SAS® software package. In Experiment 1, the weight difference between eggs produced by young and mature breeders was 10.92 g, and the component that mostly influenced this difference was the yolk (7.51 g heavier in mature breeders, compared with 4.23 g difference in albumen and 0.8 g in eggshell weights. Hatchling weight difference was 9.4 g higher in eggs from mature breeders. In Experiment 2, egg weight difference was only 0.74 g, but yolk weight was 4.59 g higher in the eggs of mature breeders. The results obtained in the present study indicate that hatchling weight is influenced by egg weight, and not by breeder age.

  10. Multi-objective Optimizations of a Novel Cryo-cooled DC Gun Based Ultra Fast Electron Diffraction Beamline

    CERN Document Server

    Gulliford, C; Bazarov, I

    2015-01-01

    We present the results of multi-objective genetic algorithm optimizations of a potential single shot ultra fast electron diffraction beamline utilizing a 225 kV dc gun with a novel cryocooled photocathode system and buncher cavity. Optimizations of the transverse projected emittance as a function of bunch charge are presented and discussed in terms of the scaling laws derived in the charge saturation limit. Additionally, optimization of the transverse coherence length as a function of final rms bunch length at sample location have been performed for three different sample radii: 50, 100, 200 microns, for two final bunch charges: 100k and 1000k electrons. Analysis of the solutions is discussed, as are the effects of disorder induced heating. In particular, a coherence length per rms spot size of 0.27 nm/micron was obtained for a final bunch charge of 100k electrons and final rms bunch length of approximately 100 fs. For a final charge of 1000k electrons the cryogun produces a coherence length per rms spot size...

  11. Hatching distribution of eggs varying in weight and breeder age

    Directory of Open Access Journals (Sweden)

    SL Vieira

    2005-06-01

    Full Text Available Broiler chicks from one incubator hatch within long periods of time, which leads to dehydration and reduction in yolk sac reserves of those chicks that have hatched earlier and potentially impairs early performance. The present research investigated the hatching distribution at intervals of incubation using eggs of different weights within one breeder age or eggs from widely different breeder ages. Eggs from breeders at 27 and 59 weeks of age (54 and 69 g and from breeders at 40 weeks of age, which were graded as light (58 g and heavy (73 g, were placed in a commercial incubator. There were a total of 1,184 eggs distributed in four treatments and eight replicates: eggs from 27-week-old breeders (27B, eggs from 59-week-old breeders (59B, light eggs from 40-week-old breeders (40BL and heavy eggs from 40-week-old breeders (40BH. Replicates were comprised of 37 eggs that were placed in each incubator tray. The treatments were physically separated from each other using a plate. Eggs were transferred to a hatcher after 432 hours of incubation and the first chick hatched at 449 hours of incubation. Afterwards, the number of completely hatched chicks from each replicate was recorded at six-hour intervals until 503 hours of incubation, when the hatchings stopped. Hatched chicks were removed from the trays after each measurement. Data were submitted to an analysis of variance with repeated measures. There was a significant interaction between breeder age and incubation length. The hatching onset of eggs from the old breeders was later compared to young breeders. Hatchability (%incubated eggs was lower for the old breeders; however, differences in hatchability as a percentage of the hatched eggs were not so evident. Complete hatchability occurred only at 503 hours of incubation; however, more than 90% eggs had hatched 18 hours earlier.

  12. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  13. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  14. Liquid-metal pumps for large-scale breeder-reactor plant (prototype pump)

    Energy Technology Data Exchange (ETDEWEB)

    Lindsay, M. (comp.)

    1976-07-01

    This report presents the recommended pump design for use in Large Scale Liquid Metal Fast Breeder Reactor plants. The base design for the pump will circulate 127,000 GPM of liquid sodium at temperatures up to 850/sup 0/F and with a total discharge head at the design point of 500 feet Na with an impeller that is 40 feet below the sodium seal. The pump design is predicated on developing an impeller design which will have a suction specific speed (S/sub n/) of about 20,000 with 20 feet NPSH available, which will result in a pump speed of 530 RPM at design conditions. The design is based on the technology developed in the design and fabrication of FFTF pumps, the design efforts for the Clinch River Breeder Reactor Pump design study and other technology.

  15. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  16. Effects of Ultra-Fast Cooling After Hot Rolling and Intercritical Treatment on Microstructure and Cryogenic Toughness of 3.5%Ni Steel

    Science.gov (United States)

    Wang, Meng; Liu, Zhenyu

    2017-07-01

    A novel process comprised of ultra-fast cooling after control rolling, intercritical quenching and tempering (UFC-LT) was applied to 3.5%Ni steel. In addition, quenching and tempering (QT) treatment was conducted in comparison. The present study focuses on the relationship between the microstructure and cryogenic toughness of 3.5%Ni steel. Results show that the microstructure of steel treated by UFC-LT consisted of tempered martensite, intercritical ferrite and two types of reversed austenite (RA) (needle shape and blocky). Compared to the QT sample, the UFC-LT sample's ultimate tensile strength decreased slightly, while its elongation increased from 32.3 to 35.7%, and its Charpy absorption energy at -135 °C increased from 112 to 237 J. The ductile-brittle transition temperature of UFC-LT sample was lower than that of the QT sample by 18 °C. The superior cryogenic toughness after UFC-LT compared to QT treatment can be attributed to the dissolution of cementite, approximately 3.0% increase in RA and the decrease in effective grain size.

  17. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  18. Vitrification by ultra-fast cooling at a low concentration of cryoprotectants in a quartz micro-capillary: a study using murine embryonic stem cells.

    Science.gov (United States)

    He, Xiaoming; Park, Eric Y H; Fowler, Alex; Yarmush, Martin L; Toner, Mehmet

    2008-06-01

    Conventional cryopreservation protocols for slow-freezing or vitrification involve cell injury due to ice formation/cell dehydration or toxicity of high cryoprotectant (CPA) concentrations, respectively. In this study, we developed a novel cryopreservation technique to achieve ultra-fast cooling rates using a quartz micro-capillary (QMC). The QMC enabled vitrification of murine embryonic stem (ES) cells using an intracellular cryoprotectant concentration in the range used for slowing freezing (1-2M). The cryoprotectants used included 2M 1,2-propanediol (PROH, cell membrane permeable) and 0.5M extracellular trehalose (cell membrane impermeable). More than 70% of the murine ES cells post-vitrification attached with respect to non-frozen control cells, and the proliferation rates of the two groups were similar. Preservation of undifferentiated properties of the pluripotent murine ES cells post-vitrification cryopreservation was verified using three different types of assays: the expression of transcription factor Oct-4, the presentation of the membrane surface glycoprotein SSEA-1, and the elevated expression of the intracellular enzyme alkaline phosphatase. These results indicate that vitrification at a low concentration (2M) of intracellular cryoprotectants is a viable and effective approach for the cryopreservation of murine embryonic stem cells.

  19. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  20. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  1. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    S. Balaguru

    2016-01-01

    Full Text Available In sodium-cooled fast reactors (SFR, grid plate is a critical component which is made of 316 L(N SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N/304 L(N SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW. This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

  2. Industrial solar breeder project using concentrator photovoltaics

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, R; Wohlgemuth, J; Burkholder, J; Levine, A; Storti, G; Wrigley, C; McKegg, A

    1979-08-01

    The purpose of this program is to demonstrate the use of a concentrating photovoltaic system to provide the energy for operating a silicon solar cell production facility, i.e., to demonstrate a solar breeder. Solarex has proposed to conduct the first real test of the solar breeder concept by building and operating a 200 kW(e) (peak) concentrating photovoltaic system based on the prototype and system design developed during Phase I. This system will provide all of the electrical and thermal energy required to operate a solar cell production line. This demonstration would be conducted at the Solarex Rockville facility, with the photovoltaic array located over the company parking lot and on an otherwise unusable flood plain. Phase I of this program included a comprehensive analysis of the application, prototype fabrication and evaluation, system design and specification, and a detailed plan for Phases II and III. A number of prototype tracking concentrator solar collectors were constructed and operated. Extensive system analysis was performed to design the Phase II system as a stand-alone power supply for a solar cell production line. Finally, a detailed system fabrication proposal for Phase II and an operation and evaluation plan for Phase III were completed. These proposals included technical, management, and cost plans for the fabrication and exercise of the proposed system.

  3. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

  4. Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining

    Energy Technology Data Exchange (ETDEWEB)

    J. A. (Bart) Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman

    2005-09-01

    The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.

  5. Parameters for quantification of hunger in broiler breeders.

    NARCIS (Netherlands)

    Jong, de I.C.; Voorst, van A.S.; Blokhuis, H.J.

    2003-01-01

    The commercial restricted feeding programme of broiler breeders has a major negative effect on welfare, as the birds are continuously hungry. Objective parameters of hunger are needed to evaluate new management or feeding systems that may alleviate hunger and thus improve broiler breeder welfare. Th

  6. Environmental Enrichment for Broiler Breeders: An Undeveloped Field

    DEFF Research Database (Denmark)

    Riber, Anja Brinch; Jong, Ingrid de; van de Werd, Heleen A.

    2017-01-01

    , cover panels, and substrate (for broiler breeders housed in cage systems). However, most of the ideas for environmental enrichment for broiler breeders need to be further developed and studied with respect to the use, the effect on behavior and welfare, and the interaction with genotype and production...

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

    Science.gov (United States)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

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

  8. The prospect of uranium nitride (UN-PuN) fuel for 25- 100MWe gas cooled fast reactor long life without refuelling

    Science.gov (United States)

    Syarifah, R. D.; Su'ud, Z.; Basar, K.; Irwanto, D.

    2016-11-01

    The prospect of uranium nitride (UN-PuN) fuel for 25-100MWe Gas Cooled Fast Reactor has been done. This research use helium coolant which has low neutron moderation, chemical inert and single phase. This study use natural uranium and plutonium. Plutonium taken from spent fuel of LWR (Light Water Reactor). So, it can reduced spent fuel in the world. The calculation use SRAC2006 and JENDL 4.0 for the data libraries. First, we calculate PIJ for fuel pin cell calculation and CITATION for core calculation. The reflector radial-axial width is 50 cm. The variation of fuel fraction is 40% until 65%, cladding 10%, and moderator 25% up to 50%. The variation of the power is 75-300 MWth (25-100 MWe). The calculation of survey parameter has been done. The variation of percentage plutonium is 7% up to 13%. We have optimum k-eff value in percentage of plutonium 11%. The high powers cause k-eff value high too. Second, the core configuration divided by three variation fuel (F1, F2, and F3). F1 is located in the central core, F2 middle core and F3 outer core. The variation percentage Plutonium for fuel F1:F2:F3 = 8%:10%:12%. The increasing power level make the burn up level increase. All case can reach burn up time plus than 20 years. The thermal powers increase cause the peak power density increase. The power 150 MWth, 225 MWth, and 300 MWth have excess reactivity (%Ak/k) less than 2%.

  9. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  10. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-11

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

  12. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    Science.gov (United States)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  13. Research and development on liquid Pb-17Li breeder in europe

    Science.gov (United States)

    Casini, G.; Sannier, J.

    1991-03-01

    Research on eutectic Pb-17Li is part of the blanket studies carried-out in the frame of the EC-Fusion Technology Programme. Two blanket concepts using liquid Pb-17Li as breeder, one water-cooled and the other self-cooled, are being investigated and are among the candidates for testing in the Next Step machines. After a brief recall of the main features of both concepts, the paper presents the progress on the Pb-17Li data base acquisition, namely: — thermophysical properties, — solubility of metallic and non metallic elements (with a special attention to tritium), — chemical reactivity, — corrosion of structural materials and related mechanical effects, — tritium production and recovery.

  14. Research and development on liquid Pb-17Li breeder in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Sannier, J. (CEA Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Inst. de Recherche Technologique et de Developpement Industriel (IRDI))

    Research on eutectic Pb-17Li is part of the blanket studies carried-out in the frame of the EC-Fusion Technology Programme. Two blanket concepts using liquid Pb-17Li as breeder, one water-cooled and the other self-cooled, are being investigated and are among the candidates for testing in the Next Step machines. After a brief recall of the main features of both concepts, the paper presents the progress on the Pb-17Li data base acquisition, namely: - thermophysical properties, - solubility of metallic and non metallic elements (with a special attention to tritium), - chemical reactivity, - corrosion of structural materials and related mechanical effects, - tritium production and recovery. (orig.).

  15. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  16. 钠冷快堆超导钠泵建模与性能研究%Modeling and Performance Research on Sodium Cooled Fast Reactor Superconductive Sodium Pump

    Institute of Scientific and Technical Information of China (English)

    杨志达; 赵佳宁; 韩伟实

    2014-01-01

    为克服钠冷快堆机械式钠泵机械磨损、噪声大、泄漏以及普通电磁泵流量小、扬程小等问题,提出了大流量鞍型磁体的电磁泵作为驱动钠循环的主泵设计方案。对其结构进行了研究并建立了相应的数学模型,利用M atlab编写程序进行了不同电流、磁感应强度和温度条件下泵的扬程、流量和效率的性能研究。结果表明,扬程随通道宽度的增加、电流的减小、磁感应强度的减小而减小,效率随电流的减小、磁感应强度的减小而减小,温度高于400℃时由于接触电阻的降低可使效率提高,鞍型超导钠泵的流量可由电流、磁感应强度控制,但较为实用的是电流控制。该研究可为具体设计提供依据。%In order to overcome the shortcoming of sodium cooled fast reactor mechanical sodium pump including mechanical wear ,large noise ,sodium leakage and the shortcom-ing of electromagnetic pump including small flow rate and small pump head ,the large flow rate sodium pump design based on saddle superconductive magnet was put forward . The structure of sodium pump was researched and the mathematic model was estab-lished .The performance including pump head ,flow rate and efficiency were researched at different currents ,magnetic flux densities and temperatures using the computer code programmed with Matlab .The result shows that the pump head will reduce with the channel width increasing ,current reducing and magnetic flux density reducing ,and the efficiency will reduce with the current reducing and magnetic flux density reducing .If the temperature overtops 400 ℃ ,the efficiency will be high because of contact resistance disappearing .T he flow rate of the saddle type superconductive sodium pump can be controlled by current and magnetic flux density ,and the current is more suitable .This research can provide foundation for concreting design .

  17. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    Energy Technology Data Exchange (ETDEWEB)

    Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

    2016-08-15

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  18. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  19. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  20. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Kondrashev, S.; Dickerson, C.; Levand, A.; Ostroumov, P. N.; Pardo, R. C.; Savard, G.; Vondrasek, R. [Physics Division, Argonne National Laboratory, Argonne, Illinois 60439 (United States); Alessi, J.; Beebe, E.; Pikin, A. [Collider-Accelerator Department, Brookhaven National Laboratory, Upton, New York 11973 (United States); Kuznetsov, G. I.; Batazova, M. A. [Budker Institute of Nuclear Physics, Novosibirsk 630090 (Russian Federation)

    2012-02-15

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  1. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Kondrashev S.; Alessi J.; Dickerson, C.; Levand, A.; Ostroumov, P.N.; Pardo, R.C.; Savard, G.; Vondrasek, R.; Beebe, E.; Pikin, A.; Kuznetsov, G.I.; Batazova, M.A.

    2012-02-03

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  2. Determination of the reactivity coefficient of a sodium cooled reactor with metallic fuel; Bestimmung der Reaktivitaetskoeffizienten eines schnellen natriumgekuehlten Reaktors mit metallischem Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Guilliard, N. [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2013-07-01

    Fast sodium cooled breeding reactors are of interest in the frame of the Generation IV reactor design. Die to the experience in France and Japan the concept seems to be realizable in the near future. Due to the new design concepts the accident scenarios and the safety analyses based on modern simulation codes have to be adjusted in the frame of the European JASMIN project. The project is aimed to develop a European accident code for fast breeder reactors based on the modular LWR code ASTEC. Extensions with respect to sodium as coolant, improved physical models and the different design are necessary. Besides this a point kinetic model shall be implemented. The coupling of point kinetic neutronics to a thermal hydraulic code requires the determination of the reactivity coefficients of the respective system. Using the core design of a benchmark specification OECD/NEA SFR task force the reactivity coefficients are determined as an example.

  3. Application Study on Ultrasonic Imaging Technique in Sodium-cooled Fast Reactor in Foreign Countries%国外超声成像技术在钠冷快堆中的应用研究

    Institute of Scientific and Technical Information of China (English)

    过明亮; 吕兆福; 段天英

    2014-01-01

    发展超声成像技术应用于钠冷快堆,对于提高反应堆安全具有重要意义。为此,国外多个国家开展了钠下超声成像技术的相关研究。本文通过对国外超声成像技术在钠冷快堆中的应用论证,以及钠回路实验台架的实验数据的介绍和研究,描述并分析了该技术在国外的研发进展和设计中的关键性因素,从而得出钠冷快堆中运用超声成像系统技术的可行性和一些亟待解决的问题。%Development of ultrasonic imaging technique used in sodium-cooled fast reactor is very important for improving the safety of the reactor.Therefore,many countries carried out study on under-sodium ultrasonic imaging technique.Based on the application of ultrasonic imaging technique to demonstrate the sodium-cooled fast reactor,and the introduction and study on the experimental data of sodium loop testing platform,this paper describes and analyzes the key factor of R&D progress abroad in this regard,and propose the feasibility of using ultrasonic imaging system technique in sodium-cooled fast reactor and some problems need to be solved.

  4. Comparative mass spectrometric analyses of Photofrin oligomers by fast atom bombardment mass spectrometry, UV and IR matrix-assisted laser desorption/ionization mass spectrometry, electrospray ionization mass spectrometry and laser desorption/jet-cooling photoionization mass spectrometry.

    Science.gov (United States)

    Siegel, M M; Tabei, K; Tsao, R; Pastel, M J; Pandey, R K; Berkenkamp, S; Hillenkamp, F; de Vries, M S

    1999-06-01

    Photofrin (porfimer sodium) is a porphyrin derivative used in the treatment of a variety of cancers by photodynamic therapy. This oligomer complex and a variety of porphyrin monomers, dimers and trimers were analyzed with five different mass spectral ionization techniques: fast atom bombardment, UV and IR matrix-assisted laser desorption/ionization, electrospray ionization, and laser desorption/jet-cooling photoionization. All five approaches resulted in very similar oligomer distributions with an average oligomer length of 2.7 +/- 0.1 porphyrin units. In addition to the Photofrin analysis, this study provides a side-by-side comparison of the spectra for the five different mass spectrometric techniques.

  5. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2014-07-01

    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  6. Effect of vaccinating breeder chickens with a killed Salmonella vaccine on Salmonella prevalences and loads in breeder and broiler chicken flocks.

    Science.gov (United States)

    Berghaus, R D; Thayer, S G; Maurer, J J; Hofacre, C L

    2011-05-01

    The objective of this study was to evaluate the effect of vaccination of breeder chickens on Salmonella prevalences and loads in breeder and broiler chicken flocks. Chickens housed on six commercial breeder farms were vaccinated with a killed Salmonella vaccine containing Salmonella Typhimurium, Salmonella Enteritidis, and Salmonella Kentucky. Unvaccinated breeders placed on six additional farms served as controls. Eggs from vaccinated and unvaccinated breeder flocks were kept separately in the hatchery, and the resulting chicks were used to populate 58 commercial broiler flock houses by using a pair-matched design. Vaccinated breeder flocks had significantly higher Salmonella-specific antibody titers than did the unvaccinated breeder flocks, although they did not differ significantly with respect to environmental Salmonella prevalences or loads. Broiler flocks that were the progeny of vaccinated breeders had significantly lower Salmonella prevalences and loads than broiler flocks that were the progeny of unvaccinated breeders. After adjusting for sample type and clustering at the farm level, the odds of detecting Salmonella in samples collected from broiler flocks originating from vaccinated breeders were 62% lower (odds ratio [95% confidence interval] = 0.38 [0.21, 0.68]) than in flocks from unvaccinated breeders. In addition, the mean load of culture-positive samples was lower in broilers from vaccinated breeders by 0.30 log most probable number per sample (95% confidence interval of -0.51, -0.09; P = 0.004), corresponding to a 50% decrease in Salmonella loads. In summary, vaccination of broiler breeder pullets increased humoral immunity in the breeders and reduced Salmonella prevalences and loads in their broiler progeny, but did not significantly decrease Salmonella in the breeder farm environment.

  7. Pre-conceptual design study on K-DEMO ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-11-15

    A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

  8. Influence of gas pressure on the effective thermal conductivity of ceramic breeder pebble beds

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Weijing [School of Civil Engineering, The University of Sydney, Sydney (Australia); Pupeschi, Simone [Institute for Applied Materials, Karlsruhe Institute of Technology (KIT) (Germany); Hanaor, Dorian [School of Civil Engineering, The University of Sydney, Sydney (Australia); Institute for Materials Science and Technologies, Technical University of Berlin (Germany); Gan, Yixiang, E-mail: yixiang.gan@sydney.edu.au [School of Civil Engineering, The University of Sydney, Sydney (Australia)

    2017-05-15

    Highlights: • This study explicitly demonstrates the influence of the gas pressure on the effective thermal conductivity of pebble beds. • The gas pressure influence is shown to correlated to the pebble size. • The effective thermal conductivity is linked to thermal-mechanical properties of pebbles and packing structure. - Abstract: Lithium ceramics have been considered as tritium breeder materials in many proposed designs of fusion breeding blankets. Heat generated in breeder pebble beds due to nuclear breeding reaction must be removed by means of actively cooled plates while generated tritiums is recovered by purge gas slowly flowing through beds. Therefore, the effective thermal conductivity of pebble beds that is one of the governing parameters determining heat transport phenomenon needs to be addressed with respect to mechanical status of beds and purge gas pressure. In this study, a numerical framework combining finite element simulation and a semi-empirical correlation of gas gap conduction is proposed to predict the effective thermal conductivity. The purge gas pressure is found to vary the effective thermal conductivity, in particular with the presence of various sized gaps in pebble beds. Random packing of pebble beds is taken into account by an approximated correlation considering the packing factor and coordination number of pebble beds. The model prediction is compared with experimental observation from different sources showing a quantitative agreement with the measurement.

  9. Investigation of V and V process for thermal fatigue issue in a sodium cooled fast reactor – Application of uncertainty quantification scheme in verification and validation with fluid-structure thermal interaction problem in T-junction piping system

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2014-11-15

    Highlights: • Outline of numerical simulation code MUGTHES for fluid-structure thermal interaction was described. • The grid convergence index (GCI) method was applied according to the ASME V and V-20 guide. • Uncertainty of MUGTHES can be successfully quantified for thermal-hydraulic problems and unsteady heat conduction problems in the structure. • Validation for fluid-structure thermal interaction problem in a T-junction piping system was well conducted. - Abstract: Thermal fatigue caused by thermal mixing phenomena is one of the most important issues in design and safety assessment of fast breeder reactors. A numerical simulation code MUGTHES consisting of two calculation modules for unsteady thermal-hydraulics analysis and unsteady heat conduction analysis in structure has been developed to predict thermal mixing phenomena and to estimate thermal response of structure under the thermal interaction between fluid and structure fields. Although verification and validation (V and V) of MUGTHES has been required, actual procedure for uncertainty quantification is not fixed yet. In order to specify an actual procedure of V and V, uncertainty quantifications with the grid convergence index (GCI) estimation according to the existing guidelines were conducted in fundamental laminar flow problems for the thermal-hydraulics analysis module, and also uncertainty for the structure heat conduction analysis module and conjugate heat transfer model was quantified in comparison with the theoretical solutions of unsteady heat conduction problems. After the verification, MUGTHES was validated for a practical fluid-structure thermal interaction problem in T-junction piping system compared with measured results of velocity and temperatures of fluid and structure. Through the numerical simulations in the verification and validation, uncertainty of the code was successfully estimated and applicability of the code to the thermal fatigue issue was confirmed.

  10. Corrosion test using commercial FMS and developed ARAA in the Experimental loop for liquid breeder

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, J. S.; Kim, S. K.; Lee, E. H.; Jin, H. G.; Lee, D. W.; Jung, Y. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, H. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A liquid-type TBM, a Helium Cooled Molten Lithium (HCML) TBM, is one of candidate TBM for KO Demos. In this concept, helium (He) and liquid lithium (Li) were used as a coolant and a breeder, respectively. However, according to our strategy for developing the liquid breeder TBM and its more relevant DEMO concept, liquid breeders not only considered liquid lithium but also lead-lithium (PbLi). An Experimental Loop for Liquid breeder (ELLI) was constructed for the purpose of validating the electromagnetic (EM) pump design, which designed and fabricated by ourselves; testing the effects of magneto-hydro-dynamics (MHD); and investigating the compatibility of PbLi using structural materials such as ferritic martensitic steel. The picture of the experimental loop is shown in Fig. 1. The performance test on each component such as heaters, the control systems for heating the loop were performed and the characteristic tests with a magnet and the EM pump were carried out. A corrosion tests using ELLI were performed with grade 91 FMS steel and ARAA during 250 hours. The corrosion tests specimens using developed commercial-scale ARAA-1 will be carried out in the ELLI loop using EM pump to investigate the corrosion behavior of ARAA-1, and the test results will be compared the previous corrosion tests of FMS corrosion specimens. A commercial FMS with grade 91 and developed ARAA steel was used for corrosion test-specimens to compare the corrosion characteristics in the flowing Pbli loop. Long term operation performance of the ELLI loop was conformed during 250 h corrosion tests. During the two separate experiments, an EM pump was operated for 250 h with a speed of 0.16 m/s (0.95 Kg/s) for corrosion tests. Mass of samples was decreased about 0.006% for ARAA-sample and 0.009% for Gr.91-sample, which corresponded to 0.22 g/m2 and 0.34 g/m2 of a corrosion attack, respectively. Unfortunately, it was unavailable to compare these weight losses with previous reports because of the

  11. 可用于小型铅铋冷快堆的核能制氢技术分析%Technical Analysis of Nuclear Hydrogen Production in Small Pb-Bi Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    孙征; 吴晓春; 李龙; 邵静

    2016-01-01

    核能制氢作为一种有前景的大规模制氢方法,得到广泛研究。该文介绍了适用于核能制氢的反应堆堆型,以及可用于核能制氢的主要方法,并对可用于小型铅铋冷快堆的核能制氢技术进行了分析。分析结果表明,小型铅铋冷快堆制氢的潜在技术路线为热化学裂解水溴钙循环或甲烷直接裂解法。%As a promising massive way,nuclear hydrogen production is being extensively investigated across the world.In this paper,the reactor types and main methods which could be used in nuclear hydrogen production were introduced,and the techniques used in Small Pb-Bi cooled fast reactor for nuclear hydrogen production were investigated.As a result,the potential technical path in Small Pb-Bi cooled fast reactor for nuclear hydrogen production were Ca-Br-Fe thermo-chemical process and methane direct pyrolysis method.

  12. Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

  13. Electron cooling

    Science.gov (United States)

    Meshkov, I.; Sidorin, A.

    2004-10-01

    The brief review of the most significant and interesting achievements in electron cooling method, which took place during last two years, is presented. The description of the electron cooling facilities-storage rings and traps being in operation or under development-is given. The applications of the electron cooling method are considered. The following modern fields of the method development are discussed: crystalline beam formation, expansion into middle and high energy electron cooling (the Fermilab Recycler Electron Cooler, the BNL cooler-recuperator, cooling with circulating electron beam, the GSI project), electron cooling in traps, antihydrogen generation, electron cooling of positrons (the LEPTA project).

  14. Electrical behaviour of ceramic breeder blankets in pebble form after γ-radiation

    Directory of Open Access Journals (Sweden)

    E. Carella

    2015-07-01

    Full Text Available Lithium orthosilicate (Li4SiO4 ceramics in from of pebble bed is the European candidate for ITER testing HCPB (Helium Cooled Pebble Bed breeding modules. The breeder function and the shielding role of this material, represent the areas upon which attention is focused. Electrical measurements are proposed for monitoring the modification created by ionizing radiation and at the same time provide information on lithium movement in this ceramic structure. The electrical tests are performed on pebbles fabricated by Spray-dryer method before and after gamma-irradiation through a 60Co source to a fluence of 4.8 Gy/s till a total dose of 5 ∗ 105 Gy. The introduction of thermal annealing treatments during the electrical impedance spectroscopy (EIS measurements points out the recombination effect of the temperature on the γ-induced defects.

  15. Stochastic Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Blaskiewicz, M.

    2011-01-01

    Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

  16. Integrated system for temperature-controlled fast protein liquid chromatography comprising improved copolymer modified beaded agarose adsorbents and a travelling cooling zone reactor arrangement.

    Science.gov (United States)

    Müller, Tobias K H; Cao, Ping; Ewert, Stephanie; Wohlgemuth, Jonas; Liu, Haiyang; Willett, Thomas C; Theodosiou, Eirini; Thomas, Owen R T; Franzreb, Matthias

    2013-04-12

    An integrated approach to temperature-controlled chromatography, involving copolymer modified agarose adsorbents and a novel travelling cooling zone reactor (TCZR) arrangement, is described. Sepharose CL6B was transformed into a thermoresponsive cation exchange adsorbent (thermoCEX) in four synthetic steps: (i) epichlorohydrin activation; (ii) amine capping; (iii) 4,4'-azobis(4-cyanovaleric acid) immobilization; and 'graft from' polymerization of poly(N-isopropylacrylamide-co-N-tert-butylacrylamide-co-acrylic acid-co-N,N'-methylenebisacrylamide). FT-IR, (1)H NMR, gravimetry and chemical assays allowed precise determination of the adsorbent's copolymer composition and loading, and identified the initial epoxy activation step as a critical determinant of 'on-support' copolymer loading, and in turn, protein binding performance. In batch binding studies with lactoferrin, thermoCEX's binding affinity and maximum adsorption capacity rose smoothly with temperature increase from 20 to 50 °C. In temperature shifting chromatography experiments employing thermoCEX in thermally jacketed columns, 44-51% of the lactoferrin adsorbed at 42 °C could be desorbed under binding conditions by cooling the column to 22 °C, but the elution peaks exhibited strong tailing. To more fully exploit the potential of thermoresponsive chromatography adsorbents, a new column arrangement, the TCZR, was developed. In TCZR chromatography, a narrow discrete cooling zone (special assembly of copper blocks and Peltier elements) is moved along a bespoke fixed-bed separation columnfilled with stationary phase. In tests with thermoCEX, it was possible to recover 65% of the lactoferrin bound at 35 °C using 8 successive movements of the cooling zone at a velocity of 0.1mm/s; over half of the recovered protein was eluted in the first peak in more concentrated form than in the feed. Intra-particle diffusion of desorbed protein out of the support pores, and the ratio between the velocities of the cooling

  17. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  18. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  19. Prevalence of Campylobacter jejuni in poultry breeder flocks

    Directory of Open Access Journals (Sweden)

    Ludovico Dipineto

    2010-01-01

    Full Text Available The aim of this work is to present the preliminary results of a study about the prevalence of Campylobacter jejuni in poultry breeder flocks. It was examined three different breeder flocks of Bojano in Molise region. A total of 360 cloacal swabs and 80 enviromental swabs was collected. Of the 3 flocks studied, 6.9% tested were positive for Campylobacter spp. The most-prevalent isolated species is C. jejuni (8.2%. Only 3 of the 360 cloacal swabs samples examined were associated with C. coli. The environmental swabs resulted negative. This results confirms again that poultry is a reservoir of this germ.

  20. Are Kirindy sifaka capital or income breeders? It depends.

    Science.gov (United States)

    Lewis, R J; Kappeler, P M

    2005-11-01

    The capital and income breeding framework has only recently been used to explain variation in female reproductive strategies in primates. The application of this framework to primates and other mammals with long reproductive cycles has not been consistent. We evaluated data on Verreaux's sifaka (Propithecus verreauxi verreauxi) in the Kirindy Forest of western Madagascar to determine whether they are capital or income breeders. We found that Verreaux's sifaka can be classified as either capital or income breeders, depending on how these concepts are operationalized. These conflicting findings highlight why the capital/income framework is currently problematic and must be standardized if it is to be a useful framework for primatologists.

  1. 可视化钠冷快堆工程信息系统的研究和设计%Viewing Sodium Cooled Fast Reactor Research and Design Engineering Information Systems

    Institute of Scientific and Technical Information of China (English)

    林全兵

    2014-01-01

    October 24, 2012, the State Council approved the "Nuclear Security Plan (2011-2020)" and "long-term nuclear power development plan (2011-2020)." This means that since the Fukushima nuclear leak in Japan since shut down the nuclear power project in China for nearly 20 months after the last full restart. Nuclear construction and re-enter the fast track scale development, on the basis of the domestic and international nuclear power project management systems, 3D visualization techniques designed to analyze the results of relevant research studies, combined with the characteristics of the sodium cooled fast reactor construction project information system application information integration technology to build a visual sodium-cooled fast reactor engineering information systems management platform to meet the use of nuclear power construction project management unit. Establish the visual content of the paper on sodium-cooled fast reactor engineering information platform has universal reference and high practical value.%2012年10月24日,国务院常务会议讨论通过了《核电安全规划(2011-2020年)》和《核电中长期发展规划(2011-2020年)》。这意味着,自日本发生福岛核泄漏以来,停摆了近20个月之后的中国核电项目终于全面重启。核电建造又重新进入了快速规模化发展的轨道,本文在对国内外核电项目管理系统、三维可视化设计技术相关研究成果进行分析研究的基础上,结合钠冷快堆工程信息系统的施工特点,应用信息集成技术,建设一个可视化的钠冷快堆工程信息系统管理平台,满足核电建造单位项目管理使用。本文的内容对可视化钠冷快堆工程信息平台的建立有普遍的借鉴意义和较高的实用价值。

  2. Development of the fast ramping 10T superconducting magnet cooled by 4K GM refrigerator; Kosoku reijigata 10T dendo reikyaku chodendo magunetto no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Kando, T.; Yoshikawa, M.; Ito, S.; Hase, T.; Hamada, M.; Hayashi, S.; Kawade, T. [Kobe Steel Ltd., Kobe (Japan); Shibuya, K.; Hirose, R. [Japan Magunet Technology Inc., Osaka (Japan)

    1999-11-10

    The ability of the small refrigerating machine is smaller than the dipping cryogenic for the conduction cooling type-superconducting magnet. Therefore, the ac loss increases, when it does the excitation speed early, and the temperature rises by the exothermic reaction and cannot be energized. The performance of the small refrigerating machine noticed that we became a refrigerating capacity of about 3W, when the thing of the 1W class is raised at 4K, and when it raises operating temperature at about 6K. We developed the conduction cooling type-superconducting magnet that it did not use oxide superconducting wire rod and superconducting cable for alternating current at the 100mm room temperature bore diameter and generates 10T in 8 minutes. (NEDO)

  3. Simulation and design of an electron beam ion source charge breeder for the californium rare isotope breeder upgrade

    Directory of Open Access Journals (Sweden)

    Clayton Dickerson

    2013-02-01

    Full Text Available An electron beam ion source (EBIS will be constructed and used to charge breed ions from the californium rare isotope breeder upgrade (CARIBU for postacceleration into the Argonne tandem linear accelerator system (ATLAS. Simulations of the EBIS charge breeder performance and the related ion transport systems are reported. Propagation of the electron beam through the EBIS was verified, and the anticipated incident power density within the electron collector was identified. The full normalized acceptance of the charge breeder with a 2 A electron beam, 0.024π  mm mrad for nominal operating parameters, was determined by simulating ion injection into the EBIS. The optics of the ion transport lines were carefully optimized to achieve well-matched ion injection, to minimize emittance growth of the injected and extracted ion beams, and to enable adequate testing of the charge bred ions prior to installation in ATLAS.

  4. CFD Analysis for Flow Behavior Characteristics in the Upper Plenum during low flow/low pressure transients for the Gas Cooled Fast Reactor (GCFR)

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Theron Marshall; Kevan Weaver; Hans Gougar

    2007-05-01

    Gas coolant at low pressure exhibits poor heat transfer characteristics. This is an area of concern for the passive response targeted by the Generation IV GCFR design. For the first 24 hour period, the decay heat removal for the GCFR design is dependent on an actively powered blower, which also would reduce the temperature in the fuel during transients, before depending on the passive operation. Natural circulation cooling initiates when the blower is stopped for the final phase of the decay heat removal, as under forced convection the core decay heat is adequately cooled by the running blower. The ability of the coolant to flow in the reverse direction or having recirculation, when the blowers are off, necessitates more understanding of the flow behavior characteristics in the upper plenum. The work done here focuses primarily on the period after the blower has been turned off, as the core is adequately cooled when the blowers are running, thus there was no need to carry out the analysis for the first 24 hours. In order to understand the plume behavior for the GCFR upper plenum several cases were run, with air, helium and helium-air mixture. For each case, the FLUENT was used to characterize the steady state velocity vectors and corresponding temperature in the upper plenum under passive decay heat removal conditions. This study will provide better insight into the plume interaction in the upper plenum at low flow and low pressure conditions.

  5. Body composition and reproduction in broiler breeders: impact of feeding strategies

    NARCIS (Netherlands)

    Emous, van R.A.

    2015-01-01

    ABSTRACT Key words: broiler breeder, feeding strategies, body composition, reproduction, behavior Nowadays, welfare issues in broiler breeders associated with nutrition and reproductive characteristics, are becoming increasingly challenging. Due to genetic selection on broilers,

  6. Factors Affecting Maternal Care in an Income Breeder, the European Roe Deer

    National Research Council Canada - National Science Library

    Reidar Andersen; Jean-Michel Gaillard; John D. C. Linnell; Patrick Duncan

    2000-01-01

    ...), which rely heavily on body reserves to raise their young. Roe deer, in contrast, are close to the income breeder end of the capital-income breeder continuum, and show high levels of maternal care. 2...

  7. Feeding broiler breeder flocks in relation to bird welfare aspects

    NARCIS (Netherlands)

    Jong, de I.C.; Krimpen, van M.M.

    2011-01-01

    To ensure health and reproductive capacity of the birds, broiler breeders are fed restricted during the rearing period, and to a lesser extent also during the production period. Although restricted feeding improves health and thereby bird welfare, on the other hand the birds are chronically hungry a

  8. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    Energy Technology Data Exchange (ETDEWEB)

    Duckett, A.J.; Sorenson, K.B.

    1985-09-01

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

  9. Big liver and spleen disease in broiler breeders in Italy

    Directory of Open Access Journals (Sweden)

    Gloria Torcoli

    2010-01-01

    Full Text Available For the first time in Italy, we have reported two outbreaks resembling big liver and spleen disease in broiler breeder flocks. The combination of clinical signs and pathological findings and the laboratory investigation results appeared corrisponde to previously recorded outbreaks in other countries.

  10. Feeding broiler breeders to improve their welfare whilst maintaining productivity

    DEFF Research Database (Denmark)

    Steenfeldt, Sanna; Nielsen, Birte Lindstrøm

    of their litter may have affected their thermoregulation. This experiment indicates that high fibre diets can alleviate the feeling of hunger currently experienced by broiler breeders, and a high ratio of insoluble fibre can reduce stereotypies and may improve the well-being of the birds....

  11. Semen bacterial flora of Rhode Island Breeder cocks in Zaria ...

    African Journals Online (AJOL)

    Semen bacterial flora of Rhode Island Breeder cocks in Zaria, Kaduna State, Nigeria. ... procedure, 27 of the 77 semen samples (35.1%) contained bacterial isolates. ... Bacteria isolates obtained from the semen include: Escherichia coli, ... destroyed in the interest of the efficient collection, preservation and delivery of highly ...

  12. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  13. Studying the capture cross sections of constructional elements from measurements of the neutron balance in breeder media

    Energy Technology Data Exchange (ETDEWEB)

    Golubev, V.I.; Dulin, V.A.; Kazanskii, Yu.A.; Darrouzet, M.; Martin-Deidier, L.; Rimpault, G.

    1987-04-01

    Until recently, the indeterminacy in the group capture cross sections of constructional elements at neutron energies above 1 keV were estimated at 15-20%, leading to an error of 0.2 and 1%, respectively, in calculating K/sub ef/ and the conversion factor of breeder reactors with oxide fuel and sodium coolant. In fact, calculations using the BNAB-78 group constants used in the USSR for the design development of fast reactors show that the mean neutron capture cross section of constructional elements (iron, nickel, and chromium) is approx. 1.4 times greater for a typical breeder reactor than in the case of calculation by the version of Carnaval IV used in France for the Superphoenix reactors. To refine the proportion of neutrons absorbed in stainless steel, the neutron balance in media consisting of uranium fuel and stainless steel with nickel in a proportion ensuring a near-unity breeder coefficient of infinite media of this composition (K/sub infinity/ = 1) has been measured at the Power-Physics Institute in Obninsk and at the Center for Nuclear Research in Caradache. The results obtained allow the accuracy of calculating the proportion of neutrons absorbed in constructional elements to be judged

  14. Ultraefficient Cooling of Resonators: Beating Sideband Cooling with Quantum Control

    Science.gov (United States)

    Wang, Xiaoting; Vinjanampathy, Sai; Strauch, Frederick; Jacobs, Kurt

    2012-02-01

    There is presently a great deal of interest in cooling high-frequency micro- and nano-mechanical oscillators to their ground states. The present state of the art in cooling mechanical resonators is a version of sideband cooling, which was originally developed in the context of cooling trapped ions. Here we present a method based on quantum control that uses the same configuration as sideband cooling--coupling the resonator to be cooled to a second microwave (or optical) auxiliary resonator--but will cool significantly colder. This is achieved by applying optimal control and varying the strength of the coupling between the two resonators over a time on the order of the period of the mechanical resonator. As part of our analysis, we also obtain a method for fast, high-fidelity quantum information transfer between resonators.

  15. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  16. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-07-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  17. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  18. Stacking with stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Caspers, Fritz E-mail: Fritz.Caspers@cern.ch; Moehl, Dieter

    2004-10-11

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles 'seen' by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly 'protected' from the Schottky noise of the stack. Vice versa the stack has to be efficiently 'shielded' against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 10{sup 5} the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters). In the 'old AA', where the antiproton collection and stacking was done in one single ring, the injected beam was further shielded during cooling by means of a movable shutter. The complexity of these systems is very high. For more modest stacking ratios, one might use azimuthal rather than radial separation of stack and injected beam. Schematically half of the circumference would be used to accept and cool new beam and the remainder to house the stack. Fast gating is then required between the high gain cooling of the injected beam and the low gain stack cooling. RF-gymnastics are used to merge the pre-cooled batch with the stack, to re-create free space for the next injection, and to capture the new batch. This scheme is less demanding for the storage ring lattice, but at the expense of some reduction in stacking rate. The talk reviews the 'radial' separation schemes and also gives some

  19. The Early Characterization of Irradiation Effects in Stainless Steels at the Experimental Breeder Reactor-II

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter

    2008-01-01

    The new Global Nuclear Energy Partnership (GNEP) program is revitalizing interest in materials development for fast spectrum reactors. With this comes the need for new, high-performance materials that are resistant to property changes caused by radiation damage. In the 1970s there was an effort to monitor the irradiation effects on stainless steels used in fast reactor cores, largely because there were a number of ‘surprises’ where materials subjected to a high flux of fast neutrons incurred dimensional and property changes that had not been expected. In the U.S., this applied to the Experimental Breeder Reactor-II. Void swelling and irradiation-induced creep caused dimensional changes in the reactor components that shortened their useful lifetime and impacted reactor operations by creating fuel handling difficulties and reactivity anomalies. The surveillance programs and early experiments studied the simplest of austenitic stainless steels, such as Types 304 and 304L stainless steel, and led to some basic understanding of the links between these irradiation effects and microchemical changes within the steel caused by operational variables such as temperature, neutron flux and neutron fluence. Some of the observations helped to define later alloy development programs designed to produce alloys that were much more resistant to the effects of neutron irradiation.

  20. Charge breeder for the SPIRAL1 upgrade: Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Maunoury, L., E-mail: maunoury@ganil.fr; Delahaye, P.; Dubois, M.; Bajeat, O.; Frigot, R.; Jeanne, A.; Jardin, P.; Kamalou, O.; Lecomte, P.; Osmond, B.; Peschard, G.; Savalle, A. [GANIL, Bd H. Becquerel BP 55027, F-14076 Caen Cedex 05 (France); Angot, J.; Sole, P.; Lamy, T. [LPSC - Université Grenoble Alpes - CNRS/IN2P3, 53 rue des Martyrs, F-38026 Grenoble Cedex (France); Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

    2016-02-15

    In the framework of the SPIRAL1 upgrade under progress at the GANIL lab, the charge breeder based on a LPSC Phoenix ECRIS, first tested at ISOLDE has been modified to benefit of the last enhancements of this device from the 1+/n+ community. The modifications mainly concern the 1 + optics, vacuum techniques, and the RF—buffer gas injection into the charge breeder. Prior to its installation in the midst of the low energy beam line of the SPIRAL1 facility, it has been decided to qualify its performances and several operation modes at the test bench of LPSC lab. This contribution shall present preliminary results of experiments conducted at LPSC concerning the 1 + to n+ conversion efficiencies for noble gases as well as for alkali elements and the corresponding transformation times.

  1. Effects of a Mixed Zone on TGO Displacement Instabilities of Thermal Barrier Coatings at High Temperature in Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Jian Wang

    2016-01-01

    Full Text Available Thermally grown oxide (TGO, commonly pure α-Al2O3, formed on protective coatings acts as an insulation barrier shielding cooled reactors from high temperatures in nuclear energy systems. Mixed zone (MZ oxide often grows at the interface between the alumina layer and top coat in thermal barrier coatings (TBCs at high temperature dwell times accompanied by the formation of alumina. The newly formed MZ destroys interface integrity and significantly affects the displacement instabilities of TGO. In this work, a finite element model based on material property changes was constructed to investigate the effects of MZ on the displacement instabilities of TGO. MZ formation was simulated by gradually changing the metal material properties into MZ upon thermal cycling. Quantitative data show that MZ formation induces an enormous stress in TGO, resulting in a sharp change of displacement compared to the alumina layer. The displacement instability increases with an increase in the MZ growth rate, growth strain, and thickness. Thus, the formation of a MZ accelerates the failure of TBCs, which is in agreement with previous experimental observations. These results provide data for the understanding of TBC failure mechanisms associated with MZ formation and of how to prolong TBC working life.

  2. Stacking with Stochastic Cooling

    CERN Document Server

    Caspers, Friedhelm

    2004-01-01

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles seen by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly protected from the Schottky noise of the stack. Vice versa the stack has to be efficiently shielded against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 105, the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters)....

  3. Haemorrhagic enteritis seroconversion in turkey breeders: field observations

    Directory of Open Access Journals (Sweden)

    Raffaella Ceruti

    2010-01-01

    Full Text Available Seroconversion to viral haemorrhagic enteritis (HE was studied in seven flocks of turkey breeders (17.974 birds in total, after 20 weeks of the onset of egg production. They showed no clinical signs, and mortality rate was normal. However, the infection caused a drop in egg production lasting about five weeks (-2.32 eggs laid during this period, but had no effect on hatching parameters.

  4. Reproduction responses of broiler-breeders to anticoccidial agents.

    Science.gov (United States)

    Jones, J E; Solis, J; Hughes, B L; Castaldo, D J; Toler, J E

    1990-01-01

    Two experiments were conducted to determine the effects of anticoccidial agents on production and reproduction of broiler breeders. In Experiment 1, nicarbazin (NCZ) was fed at 20, 50, and 100 ppm. There was no depression in egg production, egg weight, or fertility from feeding these levels. As level of NCZ increased, there was a linear decrease in hatchability. The amount of 4,4'-dinitrocarbanilide (DNC) in the egg yolks increased linearly as the levels of NCZ went up; the degree of egg-shell depigmentation was directly related to the level of NCZ fed starting at 50 ppm. Experiment 2 utilized a different strain of broiler breeders. Halofuginone (3 ppm), maduramicin (5 ppm), monensin (100 ppm), narasin (70 ppm), NCZ (125 ppm), robenidine (33 ppm), and salinomycin (60 ppm) were fed to broiler breeders at the levels listed. Only NCZ reduced egg production. Narasin induced a reduction in egg weight. Both narasin and salinomycin caused a significant drop in hatchability. Feeding NCZ also induced a rapid and more severe decrease in hatchability. Monensin was the only anticoccidial agent that reduced fertility. Halofuginone, maduramicin, and robenidine had no biologically significant effect on henday production, egg weight, hatch of fertile eggs, or shell depigmentation. Feeding NCZ at 125 ppm caused a complete bleaching of brown-shell eggs by the 3rd consecutive day of treatment; but 7 days after NCZ was withdrawn from the feed, pigmentation returned to the pretreatment level.

  5. Production aspects of broiler breeders submitted to different drinker types

    Directory of Open Access Journals (Sweden)

    LP Colvero

    2014-03-01

    Full Text Available The objective of this study was of evaluate the influence of different drinker types on the egg production, water intake, mortality, poultry litter relative humidity, egg weight, eggshell percentage, and egg specific gravity of broiler breeders. The experiment was carried out in a commercial farm with 37- to 44-wk-old broiler breeders. A randomized block experimental design, consisting of two treatments (bell or nipple drinkers with four replicates of 4.000 females each, was applied. Data were submitted to analysis of variance, and means were compared by the test of Student-Newman-Keuls at 5% significance level. Birds submitted to nipple drinkers presented lower water intake (p0.05 of drinker type on egg production or mortality. Poultry litter relative humidity was lower (p<0.05 under the nipple-drinker system. Birds drinking from bell drinkers produced heavier eggs (p<0.05 between weeks 39 and 40. Hens drinking from bell drinkers laid eggs with higher specific gravity and eggshell percentage. It was concluded that nipple drinkers can be used for broiler breeders during lay.

  6. Optical stochastic cooling in Tevatron

    CERN Document Server

    Lebedev, V

    2012-01-01

    Intrabeam scattering is the major mechanism resulting in a growth of beam emittances and fast luminosity degradation in the Tevatron. As a result in the case of optimal collider operation only about 40% of antiprotons are used to the store end and the rest are discarded. Beam cooling is the only effective remedy to increase the particle burn rate and, consequently, the luminosity. Unfortunately neither electron nor stochastic cooling can be effective at the Tevatron energy and bunch density. Thus the optical stochastic cooling (OSC) is the only promising technology capable to cool the Tevatron beam. Possible ways of such cooling implementation in the Tevatron and advances in the OSC cooling theory are discussed in this paper. The technique looks promising and potentially can double the average Tevatron luminosity without increasing its peak value and the antiproton production.

  7. Effects of breeder feeding restriction programs and incubation temperatures on progeny footpad development.

    Science.gov (United States)

    Da Costa, M J; Oviedo-Rondón, E O; Wineland, M J; Wilson, J; Montiel, E

    2014-08-01

    Footpad dermatitis begins early in life, and there is evidence of individual susceptibility. An experiment was conducted to evaluate the carryover effects of breeder feed restriction programs and incubation temperatures (TEM) on progeny footpad development at hatch, and 7 and 22 d. Cobb 500 fast feathering breeders were subjected to 2 dietary feed restriction programs during rearing: skip-a-day (SAD) and every-day feeding (EDF). At 60 wk of age, eggs from each group were collected and incubated according to 2 TEM, standard (S) eggshell temperature (38.1°C) and early-low late-high (LH). This second profile had low (36.9°C) eggshell temperature for the first 3 d, and standard temperature until the last 3 d when eggs were subjected to elevated (38.9°C) eggshell temperature. At hatch, 15 chicks from each treatment combination were sampled to obtain footpads for histological analysis. Seventy-two chicks per treatment were placed in 48 cages (6/cage), and raised to 22 d. At 7 and 22 d, 1 and 2 chickens, respectively, were sampled for footpads. The BW and group feed intake were recorded to obtain BW gain and feed conversion ratio at 7 and 21 d. Histological analysis assessed thickness and total area of stratus corneum (SC), epidermis, and dermis, and total papillae height. Data were analyzed as randomized complete block design in a 2 × 2 factorial arrangement of treatments. There was a negative effect of LH TEM on performance at both ages. An interaction effect on SC area and papillae height was observed at hatch. Additionally, SAD treatment increased thickness and area of footpad dermis. At 7 d, the SC parameters of the SAD progeny were increased. Epidermis thickness was affected by treatment interaction. Furthermore, LH TEM decreased epidermis thickness and dermis area. At 22 d, interaction effects were observed in thickness and area of SC and epidermis. Incubation S TEM increased thickness and area of dermis. It was concluded that breeder feed restriction programs

  8. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

  9. Investigation of CO{sub 2} Recovery System Design in Supercritical Carbon Dioxide Power Cycle for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Seok; Jung, Hwa-Young; Ahn, Yoonhan; Cho, Seong Kuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    These are mainly possible because the S-CO{sub 2} Brayton cycle has lower compressing work than other Brayton cycles due to its high density and low compressibility near the critical point. These attributes make easier to achieve higher turbine inlet temperature. Furthermore, the coolant chemistry control and component cooling systems are relatively simple for the S-CO{sub 2} cycle unlike the steam Rankine cycle, and therefore the total plant footprint can be greatly reduced further. However, certain amount of leakage flow is inevitable in the rotating turbo-machinery since the S-CO{sub 2} power cycle is a highly pressurized system. A computational model of critical flow in turbo-machinery seal is essential to predict the leakage flow and calculate the required total mass of working fluid in S-CO{sub 2} power system. Before designing a computational model of critical flow in turbo-machinery seal, this paper will identify what the issues are in predicting leakage flow and how these issues can be successfully addressed. Also, suitability of this solution in a large scale S-CO{sub 2} power cycle will be discussed, because this solution is for the small scale. S-CO{sub 2} power cycle has gained interest especially for the SFR application as an alternative to the conventional steam Rankine cycle, since S-CO{sub 2} power cycle can provide better performance and enhance safety. This paper discussed what the problem in leakage flow is and how to deal with this problem at present. High cavity pressure causing instability of gas foil bearing and large windage losses can be reduced by booster pump used to scavenge the gas in the rotor cavity. Also, labyrinth seals can be another good solution to decrease the rotor cavity pressure. Additionally, difference between large and small scale S-CO{sub 2} power cycle in turbo-machinery leakage is addressed. It is shown that optimization of CO{sub 2} recovery system design is more important to large scale S-CO{sub 2} power cycle. For

  10. Danish Cool

    DEFF Research Database (Denmark)

    Toft, Anne Elisabeth

    2016-01-01

    Danish Cool. Keld Helmer-Petersen, Photography and the Photobook Handout exhibition text in English and Chinese by Anne Elisabeth Toft, Curator The exhibition Danish Cool. Keld Helmer-Petersen, Photography and the Photobook presents the ground-breaking work of late Danish photographer Keld Helmer...

  11. Evaluation of US demo helium-cooled blanket options

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W. [and others

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed.

  12. Selection of Breeding Stock among Australian Purebred Dog Breeders, with Particular Emphasis on the Dam

    Directory of Open Access Journals (Sweden)

    Veronika Czerwinski

    2016-11-01

    Full Text Available Every year, thousands of purebred domestic dogs are bred by registered dog breeders. Yet, little is known about the rearing environment of these dogs, or the attitudes and priorities surrounding breeding practices of these dog breeders. The objective of this study was to explore some of the factors that dog breeders consider important for stock selection, with a particular emphasis on issues relating to the dam. Two-hundred and seventy-four Australian purebred dog breeders, covering 91 breeds across all Australian National Kennel Club breed groups, completed an online survey relating to breeding practices. Most breeders surveyed (76% reported specialising in one breed of dog, the median number of dogs and bitches per breeder was two and three respectively, and most breeders bred two litters or less a year. We identified four components, relating to the dam, that were considered important to breeders. These were defined as Maternal Care, Offspring Potential, Dam Temperament, and Dam Genetics and Health. Overall, differences were observed in attitudes and beliefs across these components, showing that there is variation according to breed/breed groups. In particular, the importance of Maternal Care varied according to dog breed group. Breeders of brachycephalic breeds tended to differ the most in relation to Offspring Potential and Dam Genetics and Health. The number of breeding dogs/bitches influenced breeding priority, especially in relation to Dam Temperament, however no effect was found relating to the number of puppies bred each year. Only 24% of breeders used their own sire for breeding. The finding that some breeders did not test for diseases relevant to their breed, such as hip dysplasia in Labrador Retrievers and German Shepherds, provides important information on the need to educate some breeders, and also buyers of purebred puppies, that screening for significant diseases should occur. Further research into the selection of breeding dams

  13. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Energy Technology Data Exchange (ETDEWEB)

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)

    2011-06-15

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  14. Optimization for Fast Zone Multilayer Fuel Assembly of Mixed Supercritical Water-Cooled Reactor%混合能谱超临界水堆快谱组件优化设计

    Institute of Scientific and Technical Information of China (English)

    杨婷; 刘晓晶; 程旭

    2011-01-01

    In order to improve the safety and sustainability of a supercritical water-cooled reactor (SCWR) core, both sub-channel and MCNP analysis were carried out to assess thermal-hydraulic and neutronic performances of the fuel assembly, which was proposed for the fast zone of a mixed-spectrum SCWR (SCWR-M). This fast zone assembly had a multilayer structure and was axially divided into several seed and blanket regions. The effects of some design parameters, I. E. Axial configuration, fuel rod diameter, pitch to diameter ratio and duct wall clearance on the thermal-hydraulic and neutronic performance of assemblies were investigated and an optimized parameter ranges were obtained.%本工作从热工水力和中子物理两方面对混合能谱超临界水堆混合谱堆芯的快谱区多层组件进行优化设计.对于轴向以再生区和裂变区交替布置的快谱组件,分别改变其轴向布置方式、燃料芯块直径、栅径比及外围燃料棒距组件盒最小距离,并分析它们对组件热工和物理性能的影响,从而得到较优的参数范围,尽可能提高混合谱超临界水堆的固有安全性和经济性.

  15. Ochratoxicosis in White Leghorn breeder hens: Production and breeding performance

    Directory of Open Access Journals (Sweden)

    Zahoor Ul Hassan*, Muhammad Zargham Khan, Ahrar Khan, Ijaz Javed1, Umer Sadique2 and Aisha Khatoon

    2012-10-01

    Full Text Available This study was designed to evaluate the effect of Ochratoxin A (OTA upon production and breeding parameters in White Leghorn (WL breeder hens. For this purpose, 84 WL breeder hens were divided into seven groups (A-G. The hens in these groups were maintained on feed contaminated with OTA @ 0.0 (control, 0.1, 0.5, 1.0, 3.0, 5.0 and 10.0 mg/Kg, respectively for 21 days. These hens were artificially inseminated with semen obtained from healthy roosters kept on OTA free feed. Egg production and their quality parameters were recorded. Fertile eggs obtained from each group were set for incubation on weekly basis. At the end of the experiment, hens in each group were killed to determined gross and microscopic lesions in different organs. OTA residue concentrations were determined in extracts of liver, kidneys and breast muscles by immunoaffinity column elution and HPLC-Fluorescent detection techniques. Feeing OTA contaminated diet resulted in a significant decrease in egg mass and egg quality parameters. Liver and kidneys showed characteristic lesions of ochratoxicosis. Residue concentration (ng/g of OTA in the hens fed 10 mg/kg OTA, was the highest in liver (26.336±1.16 followed by kidney (8.223±0.85 and were least in breast muscles (1.235±0.21. Embryonic mortalites were higher, while hatachabilites of the chicks were lower in the groups fed higher doses of OTA. Feeding OTA contaminated diets to breeder hen resulted in residues accumulation in their tissues along with significantly reduced production and breeding performance.

  16. Campylobacter epidemiology from breeders to their progeny in Eastern Spain.

    Science.gov (United States)

    Ingresa-Capaccioni, S; Jiménez-Trigos, E; Marco-Jiménez, F; Catalá, P; Vega, S; Marin, C

    2016-03-01

    While horizontal transmission is a route clearly linked to the spread of Campylobacter at the farm level, few studies support the transmission of Campylobacter spp. from breeder flocks to their offspring. Thus, the present study was carried out to investigate the possibility of vertical transmission. Breeders were monitored from the time of housing day-old chicks, then throughout the laying period (0 to 60 wk) and throughout their progeny (broiler fattening, 1 to 42 d) until slaughter. All samples were analyzed according with official method ISO 10272:2006. Results revealed that on breeder farms, Campylobacter isolation started from wk 16 and reached its peak at wk 26, with 57.0% and 93.2% of positive birds, respectively. After this point, the rate of positive birds decreased slightly to 86.0% at 60 wk. However, in broiler production all day-old chicks were found negative for Campylobacter spp, and the bacteria was first isolated at d 14 of age (5.0%), with a significant increase in detection during the fattening period with 62% of Campylobacter positive animals at the end of the production cycle. Moreover, non-positive sample was determined from environmental sources. These results could be explained because Campylobacter may be in a low concentration or in a non-culturable form, as there were several studies that successfully detected Campylobacter DNA, but failed to culture. This form can survive in the environment and infect successive flocks; consequently, further studies are needed to develop more modern, practical, cost-effective and suitable techniques for routine diagnosis. © 2016 Poultry Science Association Inc.

  17. The TMSR as actinide burner and thorium breeder

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Allibert, M.; Ghetta, V. [LPSC/IN2P3/CNRS - INPG/ENSPG - UJF, 53, avenue des Martyrs, F-38026 Grenoble Cedex (France)

    2007-07-01

    Molten Salt Reactors (MSRs) are one of the six systems retained by Generation IV as a candidate for the next generation of nuclear reactors. Molten Salt Reactor is a very attractive concept especially for the Thorium fuel cycle which allows nuclear energy production with a very low production of radio-toxic minor actinides. Studies have thus been done on the Molten Salt Breeder Reactor (MSBR) of Oak-Ridge to re-evaluate this concept. They have shown that the MSBR suffers from major drawbacks concerning for example safety and reprocessing, drawbacks incompatible with any industrial development. On the other hand, the advantages of the Thorium fuel cycle were too attractive not to look further into it. With these considerations, we have reassessed the whole concept to propose an innovative reactor called Thorium Molten Salt Reactor (TMSR). Many parametric studies of the TMSR have been carried out, correlating the core arrangement and composition, the reprocessing performances, and the salt composition. In particular, by changing the moderation ratio of the core the neutron spectrum can be modified and placed anywhere between a very thermalized neutron spectrum and a relatively fast spectrum. Even if the epithermal TMSR configurations have not been completely excluded by our calculations, our studies have shown that the reactor design where there is no graphite moderator inside the core appears to be the most promising in terms of safety coefficients, reprocessing requirements, and breeding and deployment capabilities. Larger fissile matter inventories are necessary in such a reactor configuration compared to the thermalized TMSR configurations, but the resulting deployment limitation could be solved by using transuranic elements as initial fissile load. This work is based on the coupling of a neutron transport code called MCNP with the materials evolution code REM. The former calculates the neutron flux and the reaction rates in all the cells while the latter solves

  18. Corrosion susceptibility of EUROFER97 in lithium ceramics breeders

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, Teresa, E-mail: teresa.hernandez@ciemat.es; Fernández, Pilar; Vila, Rafael

    2014-03-15

    EUROFER97 specimens were exposed in vacuum to lithium silicate pebbles at 550 °C for up to 2880 h, to evaluate its corrosion susceptibility in a simulated breeder blanket environment. The specimens and pebble bed were then analyzed and characterized by SEM-EDX, XRD, and HR-TEM. The results revealed the formation of a double chromium/iron oxide corrosion layer. HR-TEM also showed that the inner layer was amorphous, while the outer was crystalline. The amorphous layer was brittle, broke easily, and became detached from the steel.

  19. REKONFIGURASI JARING DISTRIBUSI TENAGA LISTRIK MENGGUNAKAN BREEDER GENETIC ALGORITHM (BGA

    Directory of Open Access Journals (Sweden)

    Cok. Gede Indra Partha

    2009-05-01

    Full Text Available Kebutuhan akan daya listrik saat ini semakin meningkat, seiring dengan perkembangan teknologi, cara hidup, kebutuhan dan budaya di daerah tersebut. Untuk itu keandalan dan kontinuitas pelayanan, sistem transmisidan distribusi perlu ditingkatkan untuk memperoleh pelayanan yang optimal dengan losses terendah. Padapenelitian ini digunakan metode Breeder Algoritma Genetika (BGA yang telah dikembangkan dalam optimasibeban seimbang untuk rekonfigurasi jaring distribusi tegangan menengah (JTM. Proses optimasi beban dilakukandengan cara merubah switch-switch pada penyulang (sebagai gen-gen dalam kromosom jaring distribusi sehinggadiperoleh jaring distribusi yang paling optimal. Hasil analisis menggunakan BGA menunjukkan konfigurasi baruyang optimal dengan losses terendah serta lebih cepat konvergen jika dibandingkan dengan Genetic Algorithm(GA biasa.

  20. Characteristics of tritium release behavior from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Kinjyo, T.; Nishikawa, M.; Yamashita, N.; Koyama, T.; Suematsu, K.; Fukada, S. [Graduate School of Engineering Science, Kyushu Univ., Fukuoka 812-8581 (Japan); Enoeda, M. [Naka Establishment, Japan Atomic Energy Agency, Ibaraki 319-1195 (Japan)

    2008-07-15

    A tritium release model has been developed by the present authors. The tritium release curves estimated by this tritium model give good agreement with experimental curves for Li {sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3}, Li{sub 2}ZrO{sub 3} or LiAlO{sub 2} under various purge gas conditions in our out-of-pile bred tritium release. The characteristics of tritium release behavior from various solid breeder materials carried out by us and in EXOTIC experiments at Petten are discussed in this study. (authors)

  1. A revised artificial insemination schedule for broiler breeder hens.

    Science.gov (United States)

    Van Krey, H P; Siegel, P B

    1976-03-01

    Broiler type hens maintained in individual cages were artificially inseminated with either 0.023, 0.035, or 0.047 ml. of pooled semen. Insemination intervals were nine, nine, and ten days on a repetitive basis. This schedule was adopted because it allowed an extension of the conventional insemination interval and yet remained compatible with a five-day industrial work-week. The results demonstrated that maintaining broiler breeder hens in cages and utilizing artificial insemination as a means of obtaining fertile eggs were feasible. The data also indicated that extending the insemination interval to 10 days is possible providing the number of spermatozoa inseminated is increased.

  2. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

  3. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  4. Fast neutron breeder reactor Rapsodie - situation of physics, hydraulic, thermal and dynamics studies and studies of stability early in 1963; Pile rapide rapsodie - point des etudes neutroniques, hydrauliques, thermiques et dynamiques et des etudes de stabilite au debut de l'annee 1963

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    Early in 1963, it was necessary to make a choice among the two fuels examined for Rapsodie: the UPuMo alloy with double cladding, Nb and stainless steel, and the UO{sub 2}-PuO{sub 2} mix oxide. This report presents the results of the studies effected with the two types of fuel. We reconsider at first the different models which have been studied and we give a detailed description of the alloy and oxide cores as they are envisaged early in 1963. We give then the most important physics performances of the two cores: neutron flux and spectrum, reactivity of the compensation find safety rods, neutrons balance, specific power, effective fraction of delayed neutrons, lifetime of the prompt neutrons, reactivity coefficient. We describe the hydraulic studies and experiments which have been done concerning the two cores. We discuss the criteria adopted as basis for the flow calculations. We give the results of pressure drop and sub-assembly lifting, force measurements, and vibration and pin flow distribution experiments. We discuss the constants utilized for the thermal calculations and we give the temperatures of sodium and alloy or oxide fuel, the temperature increases due to the hot points, and the limitation of the oxide fuel burn-up, originated by the pressure of the fission gases. We treat the hypotheses having been utilized for the dynamics calculations and we describe the different accidents which have been studied. We give the results of the calculations for every accident and each fuel, and we show fuel melting or sodium boiling can be avoided, even in case of the most pessimistic hypotheses, by modifying reactor characteristics (shim-rod reactivity or power of the reactor with only one cooling circuit). The reactor stability has been evaluated with the hypotheses utilized for the dynamics calculations, except of the Doppler coefficient which was intentionally increased. We show that the alloy and oxide cores are stable for every envisaged reactor power. (authors

  5. Group size adjustment to ecological demand in a cooperative breeder.

    Science.gov (United States)

    Zöttl, Markus; Frommen, Joachim G; Taborsky, Michael

    2013-04-07

    Environmental factors can determine which group size will maximize the fitness of group members. This is particularly important in cooperative breeders, where group members often serve different purposes. Experimental studies are yet lacking to check whether ecologically mediated need for help will change the propensity of dominant group members to accept immigrants. Here, we manipulated the perceived risk of predation for dominant breeders of the cooperatively breeding cichlid fish Neolamprologus pulcher to test their response to unrelated and previously unknown immigrants. Potential immigrants were more readily accepted if groups were exposed to fish predators or egg predators than to herbivorous fish or control situations lacking predation risk. Our data are consistent with both risk dilution and helping effects. Egg predators were presented before spawning, which might suggest that the fish adjust acceptance rates also to a potential future threat. Dominant group members of N. pulcher apparently consider both present and future need of help based on ecological demand. This suggests that acceptance of immigrants and, more generally, tolerance of group members on demand could be a widespread response to ecological conditions in cooperatively breeding animals.

  6. ALGERIAN SHEEP ARE NONSEASONAL BREEDERS: "CLINICAL, CYTOLOGICAL AND HISTOLOGICAL STUDIES"

    Directory of Open Access Journals (Sweden)

    A NIAR

    2001-12-01

    Full Text Available 4377 pregnant ewes have been used in this study to realise a curve of lambing. This curve prove that the three most important breeds of Algerian sheep can breed all year round.                 To prove also that Algerian sheep are no  seasonal breeder, a cytological and histological study of ewes vaginal mucosa have been done. The exfoliative cytology over all the phases of the oestrous cycle and pregnancy consistent in appearance and appear to be influenced directly by the changes in endogenous ovarian steroid hormones. The histological features of the anterior vaginal epithelium of the ewe sampled all year round, before, during, after oestrus and at pregnancy are described. In the present study, we have never found an arrest of the cyclical genital activity of ewes at any moment of the year. This finding agree with results of the curve of lambing. We can conclude that Algerian ewes are nonseasonal breeders.

  7. Helium-cooling in fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Baxi, C.B.; Hamilton, C.J.; Schleicher, R.W.; Streckert, H.

    1994-11-01

    This paper reviews different helium-cooled first wall and blanket designs; and compares the selection of structural materials. The authors found that the solid breeder, SiC-composite material option generates the lowest amount of induced radioactivity and afterheat and has the highest temperature capability. When combined with the direct cycle gas turbine system, it has the potential to be the most economical fusion system and can compete with advanced fission reactors. When compared to martensitic steel and V-alloy, SiC-composite is the least developed of these three structural materials, a focused development effort will be needed. Fundamental research has begun in addressing the issues of optimized composite materials, irradiation effects, leak tightness and low activation braze materials. Development of helium-cooled high heat flux components and further development of the direct cycle gas turbine system will also be needed.

  8. Respiratory diseases and allergic sensitization in swine breeders: a population-based cross-sectional study.

    Science.gov (United States)

    Galli, Luigina; Facchetti, Susanna; Raffetti, Elena; Donato, Francesco; D'Anna, Mauro

    2015-11-01

    The daily occupation as a swine breeder involves exposure to several bacterial components and organic dusts and inhalation of a large amount of allergens. To investigate the risk of respiratory diseases and atopy in swine breeders compared with the general population living in the same area. A population-based cross-sectional study was conducted in an agricultural area of northern Italy that enrolled a random sample of resident male breeders and non-breeders. Demographic features, comorbidities, and presence of allergic respiratory disease were retrieved through interview. Prick tests for common allergens were performed. An evaluation of pollen and mold in air samples taken inside and outside some swine confinement buildings also was performed. One hundred one male breeders (78 native-born, mean age ± SD 43.0 ± 11.1 years) and 82 non-breeders (43.0 ± 11.1 years) were enrolled. When restricting the analysis to native-born subjects, breeders vs non-breeders showed a lower prevalence of respiratory allergy (12.8% vs 31.1%, respectively, P = .002), asthma (6.4% vs 15.8%, P = .059), rhinitis (16.7% vs 51.2%, P swine buildings investigated, particularly when the pigs were inside vs outside the buildings. This study suggests that swine breeding does not increase, and might decrease, the risk of pollen sensitization and allergic disease. Copyright © 2015 American College of Allergy, Asthma & Immunology. Published by Elsevier Inc. All rights reserved.

  9. Educating the Next Generation of Plant Breeders: The Need and the Challenge

    Science.gov (United States)

    Plant breeding is critical to the future of productive agriculture, food security, and economic prosperity. Increasingly, many plant breeders are working in industry or governmental agencies that do not include education of the next generation of plant breeders as part of their mission. At the sam...

  10. Molecular characterization of chicken infectious anemia viruses detected from breeder and broiler chickens in South Korea.

    Science.gov (United States)

    Kim, H-R; Kwon, Y-K; Bae, Y-C; Oem, J-K; Lee, O-S

    2010-11-01

    In South Korea, 32 sequences of chicken infectious anemia virus (CIAV) from various flocks of breeder and commercial chickens were genetically characterized for the first time. Phylogenetic analysis of the viral protein 1 gene, including a hypervariable region of the CIAV genome, indicated that Korean CIAV strains were separated into groups II, IIIa, and IIIb. Strains were commonly identified in great-grandparent and grandparent breeder farms as well as commercial chicken farms. In the field, CIAV strains from breeder farms had no clinical effects, but commercial farm strains were associated with depression, growth retardation, and anemia regardless of the group from which the strain originated. In addition, we identified 7 CIAV genomes that were similar to vaccine strains from vaccinated and unvaccinated breeder flocks. These data suggest that further studies on pathogenicity and vaccine efficacy against the different CIAV group are needed, along with continuous CIAV surveillance and genetic analysis at breeder farms.

  11. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state...

  12. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  13. Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application%ATHLET程序的钠冷快堆应用扩展及其验证

    Institute of Scientific and Technical Information of China (English)

    周翀; Klaus Huber; 程旭

    2013-01-01

    System analysis code is important for the global simulation of the sodium-cooled fast reactor (SFR) system as well as transient and accident safety analysis .In this paper ,the best estimate system code ATHLET for light water reactors ,developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) in Germany ,was modified for SFR application .Thermal-dynamic and transport properties as well as heat transfer correlations for sodium were implemented into the ATHLET code .The modified code was then applied to simulate the Phenix reactor in France ,and validation of the code was conducted with the Phenix reactor natural convection test .The calculation results were compared with the test data .The results show that the modified ATHLET code has good applicability in simulating SFR systems .%系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构(GRS)开发的轻水堆最佳估算系统程序ATHLET 进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰(Phenix )反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。

  14. 钠冷快堆非能动余热排出系统实时仿真%Real-Time Simulation of PRHRs in Sodium-Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    张国强; 孙晓龙; 马锐; 夏庚磊

    2016-01-01

    In order to research the dynamic residual heat removal system (PRHRs ) operation characteristic of sodium cooled fast reactor,simulation model of main flow and equipment in the PRHRs were set up,using real-time two-phase multicomponent modeling tools JTopmeret,and through self-programming air heat exchanger,natural circulation module was developed,and will the coupling of thew above two,the PRHRs real-time simulation model was obtained and the running characteristics of the PRHRs simulation were researched.Simulation results show that most of the important parameters of the simulation data and design data of the relative error is within 2%,and the dynamic trend of the standby condition into the accident conditions of the system is consistent with the trend of theory.%为了研究钠冷快堆非能动余热排出系统(PRHRs)的运行特性,使用实时两相多组分建模工具 JTopmeret 搭建了 PRHRs 中主要流路与设备的仿真模型,通过自编程序开发了空气热交换器中空气自然循环模块,将二者耦合获得了 PRHRs 实时仿真模型,并对 PRHRs 的运行特性进行了仿真研究。仿真结果表明,绝大部分重要参数的仿真数据与设计数据的相对误差在2%之内,系统由备用工况进入事故工况的动态趋势与理论趋势一致。

  15. Rectlinear cooling scheme for bright muon sources

    Energy Technology Data Exchange (ETDEWEB)

    Stratakis, Diktys [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-05-03

    A fast cooling technique is described that simultaneously reduces all six phase-space dimensions of a charged particle beam. In this process, cooling is accomplished by reducing the beam momentum through ionization energy loss in absorbers and replenishing the momentum loss only in the longitudinal direction rf cavities. In this work we review its main features and describe the main results.

  16. Cool snacks

    DEFF Research Database (Denmark)

    Grunert, Klaus G; Brock, Steen; Brunsø, Karen

    2016-01-01

    such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...

  17. The Effect of Low-Density Broiler Breeder Diets on Performance and Immune Status of their Offspring

    NARCIS (Netherlands)

    Enting, H.; Boersma, W.J.A.; Cornelissen, J.B.W.J.; Winden, van S.C.L.; Verstegen, M.W.A.; Aar, van de P.J.

    2007-01-01

    Effects of low-density broiler breeder diets on offspring performance and mortality were studied using 2,100 female and 210 male Cobb 500 breeders. Breeder treatments involved 4 experimental groups and a control group with normal density diets (ND, 2,600 kcal of AME/kg during rearing and 2,800 kcal

  18. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  19. Stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Bisognano, J.; Leemann, C.

    1982-03-01

    Stochastic cooling is the damping of betatron oscillations and momentum spread of a particle beam by a feedback system. In its simplest form, a pickup electrode detects the transverse positions or momenta of particles in a storage ring, and the signal produced is amplified and applied downstream to a kicker. The time delay of the cable and electronics is designed to match the transit time of particles along the arc of the storage ring between the pickup and kicker so that an individual particle receives the amplified version of the signal it produced at the pick-up. If there were only a single particle in the ring, it is obvious that betatron oscillations and momentum offset could be damped. However, in addition to its own signal, a particle receives signals from other beam particles. In the limit of an infinite number of particles, no damping could be achieved; we have Liouville's theorem with constant density of the phase space fluid. For a finite, albeit large number of particles, there remains a residue of the single particle damping which is of practical use in accumulating low phase space density beams of particles such as antiprotons. It was the realization of this fact that led to the invention of stochastic cooling by S. van der Meer in 1968. Since its conception, stochastic cooling has been the subject of much theoretical and experimental work. The earliest experiments were performed at the ISR in 1974, with the subsequent ICE studies firmly establishing the stochastic cooling technique. This work directly led to the design and construction of the Antiproton Accumulator at CERN and the beginnings of p anti p colliding beam physics at the SPS. Experiments in stochastic cooling have been performed at Fermilab in collaboration with LBL, and a design is currently under development for a anti p accumulator for the Tevatron.

  20. Impact of radionuclides on maintenance of Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Olson, W.H.

    1985-01-01

    More than 20 years of Experimental-Breeder-Reactor-II (EBR-II) operation has demonstrated the capability to maintain radioactive equipment without undue radiation exposure to operating and maintenance personnel. The dominant radioisotopes in EBR-II primary systems are the activated corrosion product /sup 54/Mn and the fission products /sup 90/Sr and /sup 137/Cs. The presence of radioisotopes from direct activation, deposit of activated corrosion products, and release of fission products from breached fuel elements dictates special procedures, equipment, and planning but does not prohibit maintenance activities. Since 1977, the average yearly exposure of operating and maintenance personnel has been reduced while the radioactivity of systems and components has increased.

  1. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Science.gov (United States)

    Kolb, Matthias H. H.; Rolli, Rolf; Knitter, Regina

    2017-06-01

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20-30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  2. Solar-driven high temperature radiant cooling

    Institute of Scientific and Technical Information of China (English)

    SONG ZhaoPei; WANG RuZhu; ZHAI XiaoQiang

    2009-01-01

    Solar energy is widely used as one of the most important renewable energy. In addition to the growing applications of solar PV and solar water heater, solar cooling is also considered very valuable and the related researches are developing fast because of the synchronism between solar irradiance and building cooling load. Current studies mainly focus on the high temperature solar collector technique and heat-driven cooling technique, while little concern has been paid to the transport process of cooling power. In this paper, the high temperature radiant cooling is studied as an alternative way for transporting cooling power, and the performance of the combination of radiant ceiling and solar cooling is also studied. From simulation and theoretical analysis results, high temperature radiant cooling terminal shows better cooling power transportation ability against conventional air-conditioning terminal, and its thermal comfort is improved. Experiment results indicate that radiant cooling can enhance the chiller's COP (Coefficient of Performance) by 17% and cooling power regeneration by 50%.According to analysis in this paper, high temperature radiant cooling is proved to be suitable for solar cooling system, and out work can serve as a reference for later system design and promotion.

  3. Challenges for Plant Breeders from the View of Animal Nutrition

    Directory of Open Access Journals (Sweden)

    Gerhard Flachowsky

    2015-12-01

    Full Text Available The question of how to feed the growing world population is very old, but because of the increase of population and possible climate change, currently it has an explosive impact. Plant breeding can be considered as the starting point for the whole human food chain. Therefore, high, stable and highly digestible yields of phytogenic biomass with low external inputs of non-renewable resources, such as water, fuel, arable land, fertilizers, etc.; low emissions of gases with greenhouse potential during cultivation; and high resistance against biotic and abiotic stressors, including adaptation to potential climate change, and a low concentration of undesirable substances in the plants are real challenges for plant breeders in the future. Virtually unlimited resources such as sunlight, nitrogen and carbon dioxide from the air as well as the genetic pool of microbes, plants and animals can be used to breed/develop optimal plants/crops. Biofortification of plants may also be an objective of plants breeders, but it is more important for human nutrition to avoid micronutrient deficiencies. A lower concentration of undesirable substances in the plants can be considered as more important than higher concentrations of micronutrients in plants/feeds. Animal nutritionists have various possibilities for feed additive supplementation to meet animal nutrient requirements. Examples to reduce undesirable substances in feed plants are discussed and shown in the paper. In summary, plant breeding has a large and strategic potential for global feed and food security. All breeding technologies may contribute to solving important global challenges, such as sustainable use of limited global resources, improved use of unlimited resources, adaption to climate change and lowering global greenhouse gas emission. More publically supported research seems to be necessary in this field. All methods of plant breeding that contribute to a more resource-efficient production of high

  4. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H.; Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France))

    1992-01-01

    Most of the first generation of fast reactors that were operated at significant power levels employed solid metal fuels. They were constructed in the United States and United Kingdom in the 1950s and included Experimental Breeder Reactor (EBR)-I and -II operated by Argonne National Laboratory, United States, the Enrico Fermi Reactor operated by the Atomic Power Development Associates, United States and DFR operated by the U.K. Atomic Energy Authority (UKAEA). Their paper tracer pre-development of fast reactor fuel from these early days through the 1980s including ceramic fuels.

  5. Piping support system for liquid-metal fast-breeder reactor

    Science.gov (United States)

    Brussalis, Jr., William G.

    1984-01-01

    A pipe support consisting of a rigid link pivotally attached to a pipe and an anchor, adapted to generate stress or strain in the link and pipe due to pipe thermal movement, which stress or strain can oppose further pipe movement and generally provides pipe support. The pipe support can be used in multiple combinations with other pipe supports to form a support system. This support system is most useful in applications in which the pipe is normally operated at a constant elevated or depressed temperature such that desired stress or strain can be planned in advance of pipe and support installation. The support system is therefore especially useful in steam stations and in refrigeration equipment.

  6. Feasibility studies for production of {sup 89}Sr in the Fast Breeder Test Reactor (FBTR)

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Debasish; Vithya, J.; Ashok Kumar, G.V.S.; Swaminathan, K.; Kumar, R.; Venkata Subramani, C.R.; Vasudeva Rao, P.R. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam (India). Fuel Chemistry Div.

    2013-07-01

    {sup 89}Sr, a pure beta emitter with half life of 50.53 d is used as its chloride solution for palliative care of bone metastases. This paper describes the feasibility studies that have been conducted at FBTR, IGCAR for production of this radionuclide using the {sup 89}Y(n, p){sup 89}Sr reaction. Yttria pellets were irradiated in a special subassembly at the core centre for a total of 73 d in two steps of 35 d and 38 d with a time gap of 38 d. The irradiated yttria target was dissolved in nitric acid and the bulk Y was separated by solvent extraction using the TBP-HNO{sub 3} complex. The {sup 89}Sr fraction was purified using the cation exchange resin DOWEX 50W x 8 (100-200 mesh size) from the other radioactive impurities seen. The eluted {sup 89}Sr fraction was assayed using a GM counting system. The {sup 89}Sr activity produced in 1 g of yttria pellet was found to be 19 mCi. (orig.)

  7. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  8. Considerations on a critical experiment program for a large fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-09-15

    The design studies for large LMFBR in Japan are being continued by PNC and Electric Companies. Both of them have adopted loop-type and 1000 MWe class reactor as the reference design, but main parameters of the core have not yet been fixed. The main core parameters of the present design are shown. Comparing the geometrical properties with those in ZPPR-9 and 10, some of the degrees of mockup are not satisfactory. In addition, there is another difference between a reactor and a mockup critical assembly. Therefore extrapolation is important to apply the results of JUPITER to the core design.

  9. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  10. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  11. Oil Analysis by Fast DSC

    NARCIS (Netherlands)

    Wetten, I.A.; Herwaarden, A.W.; Splinter, R.; Ruth, van S.M.

    2014-01-01

    Thermal analysis of Olive and Sunflower Oil is done by Fast DSC to evaluate its potential to replace DSC for adulteration detection. DSC measurements take hours, Fast DSC minutes. Peak temperatures of the crystallisation peak in cooling for different Olive and Sunflower Oils are both comparable to D

  12. Oil Analysis by Fast DSC

    NARCIS (Netherlands)

    Wetten, I.A.; Herwaarden, A.W.; Splinter, R.; Ruth, van S.M.

    2014-01-01

    Thermal analysis of Olive and Sunflower Oil is done by Fast DSC to evaluate its potential to replace DSC for adulteration detection. DSC measurements take hours, Fast DSC minutes. Peak temperatures of the crystallisation peak in cooling for different Olive and Sunflower Oils are both comparable to

  13. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    Science.gov (United States)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  14. Development of a New Thermochemical and Electrolytic Hybrid Hydrogen Production System for Sodium Cooled FBR

    Science.gov (United States)

    Nakagiri, Toshio; Kase, Takeshi; Kato, Shoichi; Aoto, Kazumi

    A new thermo-chemical and electrolytic hybrid hydrogen production system in lower temperature range is newly proposed by the Japan Nuclear Cycle Development Institute (JAEA) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeder Reactor (FBR). The system is based on sulfuric acid (H2SO4) synthesis and decomposition process developed earlier (Westinghouse process), and sulfur trioxide (SO3) decomposition process is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce the operation temperature 200-300°C lower than Westinghouse process. SO3 decomposition with the voltage lower than 0.5V was confirmed in the temperature range of 500 to 600°C and theoretical thermal efficiency of the system evaluated based on chemical reactions was within the range of 35% to 55% under the influence of H2SO4 concentration and heat recovery. Furthermore, hydrogen production experiments to substantiate the whole process were performed. Stable hydrogen and oxygen production were observed in the experiments, and maximum duration of the experiments was about 5 hours.

  15. Cooling technique

    Energy Technology Data Exchange (ETDEWEB)

    Salamon, Todd R; Vyas, Brijesh; Kota, Krishna; Simon, Elina

    2017-01-31

    An apparatus and a method are provided. Use is made of a wick structure configured to receive a liquid and generate vapor in when such wick structure is heated by heat transferred from heat sources to be cooled off. A vapor channel is provided configured to receive the vapor generated and direct said vapor away from the wick structure. In some embodiments, heat conductors are used to transfer the heat from the heat sources to the liquid in the wick structure.

  16. Ultra-Efficient Cooling of Resonators: Beating Sideband Cooling with Quantum Control

    CERN Document Server

    Wang, Xiaoting; Strauch, Frederick W; Jacobs, Kurt

    2011-01-01

    The present state-of-the-art in cooling mechanical resonators is a version of "sideband" cooling. Here we present a method that uses the same configuration as sideband cooling --- coupling the resonator to be cooled to a second microwave (or optical) auxiliary resonator --- but will cool significantly colder. This is achieved by varying the strength of the coupling between the two resonators over a time on the order of the period of the mechanical resonator. As part of our analysis, we also obtain a method for fast, high-fidelity quantum information-transfer between resonators.

  17. 钠冷快堆燃料组件热工水力特性数值模拟与分析%Numerical Simulation and Analysis on Thermal-hydraulic Behavior of Fuel Assembly for Sodium-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 周志伟

    2014-01-01

    The thermal-hydraulic behavior of triangular arranged fuel bundle with wrapped wire spacer of fuel assembly for sodium-cooled fast reactor was investigated by employing CFD code CFX ,and the results were compared with subchannel analysis code SuperEnergy .Fuel bundles composed of 7,19 ,37 and 61 fuel rods were analyzed sepa-rately .The axial velocity ,cross flow mixing effect ,and temperature rise along axial direction for different subchannels of the fuel bundle were discussed ,and the effect of wrapped wire spacer was carefully investigated .The results show that the wrapped wire spacer plays an important role on the cross flow effect and axial velocity distribution as well as the temperature rise in different subchannels .Moreover ,with the increase of fuel rods ,the flow in fuel bundle becomes more complicated ,and the non-uniformity of the axial flow also shows a tendency to enhance .%利用CFD程序CFX ,分别对7、19、37、61根棒组成的三角形排列螺旋绕丝定位的钠冷快堆燃料组件棒束通道进行了热工水力特性的分析研究,并将结果与子通道程序SuperEnergy进行了对比验证。重点考察了棒束通道轴向流动分布、横向流交混效应及子通道轴向温升,分析了定位绕丝的影响。结果表明,绕丝对棒束通道的横向流交混效应、轴向流动分布及子通道温升有着重要影响,且随棒束的增多,通道内的流动趋向复杂化,轴向流动不均匀性有升高趋势。

  18. 加速器驱动的次临界10MW气冷快堆物理方案研究%Physical Scheme Study of 10 MW Accelerator-driven Sub-critical Gas-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    秦长平; 顾龙; 李金阳

    2013-01-01

    本文采用MCNPX与ORIGEN耦合的COUPLE2.0程序对加速器驱动的次临界气冷快堆进行了物理方案的设计和研究。计算得到了该方案350d燃耗期间的kef、质子效率、加速器束流强度、1500K温度下的多普勒系数、功率峰因子等参数,并研究了该方案的安全特性;通过计算得到了该方案燃耗并分析了该方案的嬗变能力。结果表明,该方案在350d燃耗期间的kef、加速器束流强度、功率峰因子变化较小,在假想事故下仍保持较深的次临界状态,系统安全性能较好;燃耗较浅,嬗变支持比为20.28,具有较好的嬗变效果。%A physical scheme design of a 10MW accelerator-driven sub-critical gas-cooled fast reactor was studied .The coupling program COUPLE2.0 was applied ,which couples MCNPX and ORIGEN .The physical parameters such as kef ,proton efficiency , accelerator current ,power peak factor and Doppler coefficient of 1500K varying with burnup time were obtained ,and the safety feature of the system was analyzed .By further calculation the 350 d burnup of the system was obtained and the transmutation capability was analyzed . It is found that during the 350 d burnup , kef , accelerator current ,power peak factor variations are relatively small .The system has good safety feature with deep sub-critical status in hypothetical accident . The total burnup is relatively shallow , but the system has good transmutation capability with the transmutation support ratio of 20.28 .

  19. Study of Physical Scheme for 10 MW Accelerator-driven Fast-thermal Coupled Gas-cooled Reactor%加速器驱动10 MW快热耦合气冷堆物理方案研究

    Institute of Scientific and Technical Information of China (English)

    李金阳; 顾龙; 秦长平; 王大伟; 刘璐

    2013-01-01

    The accelerator-driven sub-critical system has promising future in transmuta-tion of nuclear spent fuels .A physical design of a 10MW fast-thermal spectrum gas-cooled reactor was studied .The program COUPLE2.0 ,which couples with MCNPX and ORIGEN ,was applied to compute this scheme ,and the continuous energy neutron cross section was obtained by the nuclear data library of ENDF-7 which was modified with five different temperatures :300 ,600 ,900 ,1 500 ,and 2500K .The physical pa-rameters such as kef ,proton efficiency ,delayed neutron fraction and accelerator current varying with the burnup time were obtained .The system during 350 d burnup was obtained and the transmutation capability was analyzed by the further calculation .It is found that during the 350 d burnup ,the variations of kef and accelerator current are relatively small .The system has good transmutation capability with the transmutation support ratio of 24.86 .%加速器驱动的次临界系统(ADS )在实现嬗变核废料方面具有良好的前景。对加速器驱动10 M W次临界快热耦合能谱气冷堆的物理方案进行了设计和研究,利用 MCNPX和ORIGEN耦合的计算程序COUPLE2.0对该方案进行了计算,其中,中子截面采用ENDF-7处理后得到的5个温度300、600、900、1500、2500 K下的连续能量核数据库。得到该方案350 d燃耗期间的 kef 、质子效率、缓发中子份额以及加速器束流强度的变化。进而计算得到了该方案的燃耗信息,并分析了该方案的嬗变能力。结果表明,该方案在350 d燃耗期间的 kef 、加速器束流强度变化较小,嬗变支持比为24.86,具有较好的嬗变效果。

  20. Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Annabattula, R. [Indian Institute of Technology Madras (IITM), Department of Mechanical Engineering (India); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2014-10-15

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li{sub 4}SiO{sub 4}) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li{sub 4}SiO{sub 4} pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li{sub 4}SiO{sub 4} pebble bed, and (3) in situ measurement of thermal conductivity of the Li{sub 4}SiO{sub 4} pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses

  1. Beam Dynamics With Electron Cooling

    CERN Document Server

    Uesugi, T; Noda, K; Shibuya, S; Syresin, E M

    2004-01-01

    Electron cooling experiments have been carried out at HIMAC in order to develop new technologies in heavy-ion therapy and related researches. The cool-stacking method, in particular, has been studied to increase the intensity of heavy-ions. The maximum stack intensity was 2 mA, above which a fast ion losses occurred simulatneously with the vertical coherent oscillations. The instability depends on the working point, the stacked ion-density and the electron-beam density. The instability was suppressed by reducing the peak ion-density with RF-knockout heating.

  2. Effects of breeder age and pre-incubation storage of eggs on ...

    African Journals Online (AJOL)

    S..

    hatched chicks was influenced by the storage period x breeder age ... liver weight as a percentage of live weight was higher when eggs obtained from 10 ... and it remains unclear how these storage conditions affect water vapour loss during.

  3. ATLAS - Liquid Cooling Systems

    CERN Multimedia

    Bonneau, P.

    1998-01-01

    Photo 1 - Cooling Unit - Side View Photo 2 - Cooling Unit - Detail Manifolds Photo 3 - Cooling Unit - Rear View Photo 4 - Cooling Unit - Detail Pump, Heater and Exchanger Photo 5 - Cooling Unit - Detail Pump and Fridge Photo 6 - Cooling Unit - Front View

  4. Capillariosis in breeder discus (Symphysodon aequifasciatus in Iran

    Directory of Open Access Journals (Sweden)

    Rahmati-Holasoo Hooman

    2010-01-01

    Full Text Available The global ornamental fish trade is a rapidly growing industry. Cultivation and propagation of ornamental fishes have been increasing in the last 20 years in Iran. Discus (Symphysodon aequifasciatus from Cichlidae is one of the most popular and expensive aquarium fish. In the past few years farming of this fish has been well developed in Iran. Two breeder discus fish (Symphysodon aequifasciatus from two different propagation centres (with high mortality with signs of anorexia, loss of balance, moribundity and darkness in skin colour were referred to Laboratory of Aquatic Diseases of Veterinary Faculty, University of Tehran. After the survey of ectoparasites, necropsy was performed under aseptic conditions; bacterial culture on standard media was done and the alimentary canal was extruded. In both fish no ectoparasite was detected and no bacteria from these cases grew on the standard media. In internal survey 5 and 25 nematodes were detected in each fish. A high number of free eggs were observed in intestine of fish. Regarding morphological characteristics of the nematodes and their eggs, they were identified as Capillaria sp. Treatment of other fish with levamisole was effective and the loss was terminated. Some helminthes like Capillaria pterophylli Heinze, 1933, can cause a high mortality in cichlid aquarium fishes. This study showed that infection with some species of Capillaria could cause a heavy loss in ornamental fish from Cichlidae. Diagnosis of parasites of these fishes can help us to prevent high mortalities.

  5. Bacterial orchitis and epididymo-orchitis in broiler breeders.

    Science.gov (United States)

    Monleon, Rafael; Martin, Michael P; John Barnes, H

    2008-12-01

    Two cases of orchitis and epididymo-orchitis in broiler breeders are described. The first case occurred in a 62-week-old rooster showing signs of breeding inactivity. Grossly, the left testis was abnormally shaped and had multiple yellow foci, prominent blood vessels, and a gelatinous mass on the surface. The right testis and other tissues appeared normal. A pure, heavy growth of Staphylococcus aureus was obtained on bacterial culture, and intralesional Gram-positive cocci were numerous on histopathology. The second case occurred in a 28-week-old rooster that was found dead. Both testes were abnormally shaped, swollen, firm, and had irregular, dark, depressed areas, prominent vessels, and multiple petechial haemorrhages. Epididymides were enlarged and nodular. Other lesions in this rooster included poor nutritional condition, distended cloaca, and urate scalding of the abdominal skin below the vent. Microscopically, there was extensive, severe heterophilic intratubular orchitis and epididymitis with intralesional Gram-negative bacteria. A heavy, pure growth of Escherichia coli was obtained from the testes and epididymides. In both cases, the findings suggested that the most probable route of infection was ascending via the ductus deferens.

  6. Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

    Energy Technology Data Exchange (ETDEWEB)

    Thirumalai, M., E-mail: mtl@igcar.gov.i [Fast Reactor Technology Group, Department of Atomic Energy, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India); Anandaraj, M.; Kumar, P. Anup [Fast Reactor Technology Group, Department of Atomic Energy, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India); Prakash, V., E-mail: prakash@igcar.gov.i [Fast Reactor Technology Group, Department of Atomic Energy, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India); Anandbabu, C.; Kalyanasundaram, P.; Vaidyanathan, G. [Fast Reactor Technology Group, Department of Atomic Energy, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

    2010-01-15

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m{sup 3}/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.

  7. Digestible Threonine Levels in the Starter Diet of Broilers Derived from Breeders of Different Ages

    Directory of Open Access Journals (Sweden)

    CBGS Tanure

    2015-12-01

    Full Text Available ABSTRACT The aim of this study was to evaluate the effect of digestible threonine supplementation in the starter diet on the performance, intestinal parameters, and nutrient metabolism of broilers derived from breeders of different ages. In total, 480 one-day-old Cobb chicks, derived from 38-or 49-week-oldbreeders, were housed in experimental battery cages until 21 days of age and fed four different threonine levels (800, 900, 1,000, or 1,100 mg/kg in the starter feed. A completely randomized experimental design in a 2x4 factorial arrangement (breeder age x threonine levels was applied, totaling eight treatments with five replicates of 12 birds each. Broilers from older breeders fed 800 mg digestible threonine/kg of diet presented higher weight gain, with a positive linear effect. There was also an interaction between breeder age and threonine levels for the weight gain of 21-d-old broilers supplemented at maximum level of 1,003 mg Thr/kg diet during the starter phase. There was no effect of breeder age or threonine levels on nutrient metabolism during the period of 17-21 days. There was no influence of breeder age or threonine levels in the starter diet on intestinal morphometric measurements, absorption area, or percentage of goblet cells.

  8. The prevalence of subclinical endometritis and intrauterine infections in repeat breeder cows.

    Science.gov (United States)

    Pothmann, H; Prunner, I; Wagener, K; Jaureguiberry, M; de la Sota, R L; Erber, R; Aurich, C; Ehling-Schulz, M; Drillich, M

    2015-05-01

    The objectives of this study were to assess the prevalence of subclinical endometritis and the presence of common uterine pathogens in repeat breeder cows. A total of 121 cows with three or more consecutive artificial inseminations without conception and no clinical signs of disease were defined as repeat breeder cows and were enrolled in this trial. Intrauterine samples were collected with the cytobrush technique to determine the prevalence of subclinical endometritis and bacteriologic infections. Blood samples were analyzed for concentrations of progesterone and estradiol in plasma to assess ovarian activity. Furthermore, breed, parity, history of calving and postpartum uterine infection, clinical findings of transrectal palpation, and backfat thickness were analyzed as potential factors for the prevalence of subclinical endometritis in repeat breeder cows. The prevalence of subclinical endometritis in repeat breeder cows was 12.7%; but common uterine pathogens, Escherichia coli and Trueperella pyogenes, were found in only one and three cows, respectively. Ovarian activity was determined in 95.0% of all cows. Recorded variables had no effect on the prevalence of subclinical endometritis in repeat breeder cows. In conclusion, subclinical endometritis and uterine infections linked to common pathogens were playing a minor role as a cause for repeat breeder cows in this study. Alternative reasons for failure to conceive in these cows are discussed.

  9. Egg traits and physiological neonatal chick parameters from broiler breeder at different ages

    Directory of Open Access Journals (Sweden)

    BC Luquetti

    2004-03-01

    Full Text Available This investigation was carried out to study egg traits (weight, specific gravity and eggshell thickness, and neonatal chick parameters (weight, blood, cardiac and lung parameters from breeder at different ages. After hatching, neonatal chicks, male and female, from broiler breeder at three different ages (30, 45 and 60 weeks were sacrificed and blood (red blood cell number, hematocrit, mean cell volume, blood viscosity, and haemoglobin, cardiac (right ventricle and total ventricle weights, cardiac index and lung (mean pulmonary pressure, fresh relative lung weight parameters were measured. No significant differences in eggshell thickness or specific gravity were observed in eggs from the three different breeder age. The incubated eggs and neonatal chicks showed heavier absolute weights with increasing breeder age. Broiler breeder age did not affect blood parameters or cardiac index, but affected right ventricle and total ventricle absolute weights. Red blood cell number and hematocrit were higher and lung weight and lung weight:chick weight index were lower for female neonatal chicks. The findings of this study revealed that breeder age affects neonatal chick parameters such as body weight, heart and lung absolute weights, but not blood parameters.

  10. Enhancement of laser cooling by the use of magnetic gradients

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, A; Retzker, A; Plenio, M B [Institut fuer Theoretische Physik, Universitaet Ulm, 89069 Ulm (Germany); Wunderlich, C, E-mail: andreas.albrecht@uni-ulm.de [Fachbereich Physik, Universitaet Siegen, D-57068 Siegen (Germany)

    2011-03-15

    We present a laser cooling scheme for trapped ions and atoms using a combination of laser couplings and a magnetic gradient field. In a Schrieffer-Wolff transformed picture, this setup cancels the carrier and blue sideband terms completely (up to first order in the Lamb-Dicke parameter), resulting in an improved cooling behaviour compared to standard cooling schemes in the Lamb-Dicke regime (e.g. sideband cooling) and allowing cooling to the vibrational ground state. A condition for optimal cooling rates is presented and the cooling behaviour for different Lamb-Dicke parameters and spontaneous decay rates is discussed. Cooling rates of one order of magnitude less than the trapping frequency are achieved using the new cooling method. Furthermore, the scheme exhibits fast rates and low final populations, even for significant deviations from the optimal parameters, and provides good cooling rates also in the multi-particle case.

  11. Selective breeding of Arabian and Thoroughbred racehorses in Algeria: perceptions, objectives and practices of owners-breeders

    Directory of Open Access Journals (Sweden)

    Safia Tennah

    2014-04-01

    Full Text Available This survey, conducted with 461 racehorse owners-breeders in Algeria between 2009 and 2011, investigates their perceptions, objectives and practices regarding selective breeding. Racehorse breeding is a full-time professional activity for a third of interviewees. The holdings are small-sized with 77% owning one or two mares. The regular practice of mating is here used to categorize breeders according to their degree of professionalization (38.4% professional vs. 61.6% occasional breeders. Experience in the sector was also used to classify breeders, considering as "junior" the breeders under 10 years experience (38.8% and as "senior" those above 10 years (61.2%. More than professionalization, experience shows a significant impact on practices and objectives. Thus, experience influences breed choice (junior breeders tend to specialize while senior own both Arabian and Thoroughbreds, age at first foaling (sooner among senior breeders, information sources considered for selecting stallions (senior use more diversified sources, the importance granted to the price of mating (greater for junior breeders, the importance granted to the ranking compared to earnings (the ranking being more important to junior breeders, and the priority given to breeding (junior breeders give higher priority to a buy-race-resell activity. Finally, racehorse breeding is poorly professionalized, the only financial goal being cost coverage. Despite inappropriate practices, an interest for selection is noticed.

  12. Broadband optical cooling of molecular rotors

    CERN Document Server

    Lien, Chien-Yu; Odom, Brian C

    2014-01-01

    Contrary to intuition, resonant laser excitation of bound electrons can decrease the temperature of a system, with electronic relaxation times as fast as nanoseconds allowing for rapid cooling to far below ambient temperature. Although laser cooling of atoms is routine owing to their relatively simple internal structure, laser cooling of molecular translational speeds, vibrations, or rotations is challenging because a different laser frequency is required to electronically excite each populated vibrational and rotational state. Here, we show that molecules with decoupled vibrational and electronic modes can be rotationally cooled using a single spectrally filtered broadband laser to simultaneously address many rotational states. We optically cool AlH$^+$ ions held in a room-temperature radiofrequency Paul trap to collect 96% of the population in the ground quantum state, corresponding to a rotational temperature of 4 K. In our current implementation, parity-preserving electronic cycling cools to the two lowes...

  13. Cool visitors

    CERN Multimedia

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  14. Research on Test Unit for Determining Cooling Characteristics of Quenchants

    Institute of Scientific and Technical Information of China (English)

    LI Lin-lin; ZHANG Wei-min; CHEN Nai-lu; LIU Zhan-cang; LIU Yang

    2004-01-01

    According to ASTM Designation DXXXX Standard Test Method for Determination of Cooling Characteristics of Quenchants by Cooling Curve Analysis with Agitation (Drayton Unit), a test unit was designed using which the cooling characteristics of a quenchant at different flow rates and different temperatures can be determined. We discussed the effects of flow rate and temperature on cooling characteristics of fast speed quenching oil. The results show that this unit can be well used for determining cooling characteristics of quechants at different flow rates and different temperatures. At the same time our investigation can provide a reference to the establishment of a standard for determination of cooling characteristics of quenchants with agitation.

  15. A numerical investigation of the sCO{sub 2} recompression cycle off-design behaviour, coupled to a sodium cooled fast reactor, for seasonal variation in the heat sink temperature

    Energy Technology Data Exchange (ETDEWEB)

    Floyd, J., E-mail: jeremy.floyd@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Moisseytsev, A., E-mail: amoissey@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Haubensack, D., E-mail: david.haubensack@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Rodriguez, G., E-mail: gilles.rodriguez@cea.fr [CEA, DEN, Département de Technologie Nucléaire, F-13108 Saint Paul Lez Durance (France); Sienicki, J., E-mail: sienicki@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Avakian, G., E-mail: gilles.avakian@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France)

    2013-07-15

    Highlights: • Year-round behaviour of the supercritical CO{sub 2} recompression cycle is simulated. • Behaviour of the system was uncertain due to large changes in the fluid properties. • Cycle thermodynamic optimisation and component preliminary designs were performed. • No off design cycle stability issues, compressors operate away from surge region. • Independent speed control of compressors maintains power and cycle efficiency. -- Abstract: Supercritical CO{sub 2} cycles are particularly attractive for Generation IV Sodium-Cooled Fast Reactors (SFRs) as they can be simple and compact, but still offer steam-cycle equivalent efficiency while also removing potential for Na/H{sub 2}O reactions. However, CO{sub 2} thermophysical properties are very sensitive close to the critical point which raises, in particular, questions about the compressor and so cycle off-design behaviour when subject to inevitable temperature increases that result from seasonal variations in the heat sink temperature. This publication reports the numerical investigation of such an issue that has been performed using the Plant Dynamics Code (ANL, USA), the cycle being optimised for the next French SFR, ASTRID (1500 MW{sub th}), as a test-case. On design, the net plant efficiency is 42.2% for a high pressure (25 MPa) turbine with an inlet temperature of 515 °C and considering a cycle low temperature of 35 °C. The off-design cycle behaviour is studied based on preliminary designs for the main components and assuming the use of a fixed heat sink flow rate. First results obtained using a common fixed shaft speed for all turbomachines, without any other active control, show no stability issues and roughly constant density (and volumetric flow rate) at the main compressor inlet for the range of heat sink temperature considered (21–40 °C). This occurs because the new stationary states are found without requiring a significant shift of mass to the higher pressure level, meaning the

  16. 改进型快谱超临界水冷堆增殖特性初步研究%Primary Study on Breeding Property of Improved Supercritical Water Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘紫静; 于涛; 谢金森

    2012-01-01

    In this paper, the core mode of improved supercritical water cooled fast reactor is established. At first, reasonable fuel assembly design is obtained by studying the influences of seed fuel pin diameter and blanket coolant channel diameter to conversion ratio (Cr). Then, viod reactivity coefficient and CR of six different core arrangements are calculated. Finaly, the influences of fuel components to CR and void reactivity coefficient are analysed. The results show that negative void reactivity coefficient can be satisfied and Cr can be increased by reducing Hydrogen to Heavy-metal ratio (H/HM), increasing blanket assembly numbers by proper distribution. Cr is substantially increased and more negative void reactivity coefficient can be met by reducing PuO2 mass ratio in fuel, when PuO2 mass ratio reach 20.8% in MOX fuel and 235U enriched at 0.2% in UO2 fuel have been adopted as seed and blanket assmbly respectively, the sixth core program reaches CR=1.04395 and give negative void reactivity coefficient, which meets the primary requirements for SCFR breeding.%建立改进型快谱超临界水冷堆( SCFR-M)堆芯模型,探讨点火区燃料棒直径和增殖区水棒直径对堆芯转换比的影响,得到合理的燃料组件设计形式.设计计算6种不同堆芯布置下的增殖特性和空泡反应性,分析燃料组分对堆芯转换比和空泡反应性系数的影响.结果表明:减小氢原子数与重金属原子数之比(H/HM),增加堆芯增殖组件数目并采用合理布置可在满足负空泡反应系数的同时提高转换比;降低燃料中PuO2质量分数可以使转换比大幅增加,同时使堆芯的空泡反应性系数有更大负值;当点火组件采用PuO2质量分数为20.8%的MOX燃料,增殖组件采用235U富集度为0.2%的UO2燃料,方案6的设计可以使堆芯的初始转换比达到1.04395,并且空泡反应性系数为负,初步达到快谱超临界水冷堆的增殖要求.

  17. Comparison of serum leptin, glucose, total cholesterol and total protein levels in fertile and repeat breeder cows

    Directory of Open Access Journals (Sweden)

    Saime Guzel

    2014-12-01

    Full Text Available In the present study we measured serum glucose, leptin, total cholesterol and total protein concentrations in repeat breeder cows and compared them with fertile cows. For this aim, 20 repeat breeder cows and 20 fertile cows were used as material. Repeat breeder cows were found to have lower levels of leptin and glucose as compared with fertile ones. No significant differences in total cholesterol and total protein levels were observed between the two groups. No significant correlation of leptin with glucose, total cholesterol and total protein was observed in fertile and repeat breeder cows. Low concentrations of glucose and leptin can have some effects on reproductive problems as repeat breeder and help to understand potential mechanisms impairing fertility in repeat breeder cows.

  18. Effects of breeder hen age and dietary L-carnitine on progeny embryogenesis.

    Science.gov (United States)

    Peebles, E D; Kidd, M T; McDaniel, C D; Tanksley, J P; Parker, H M; Corzo, A; Woodworth, J C

    2007-06-01

    1. Ross 308 broiler breeder hens were given diets containing 0 or 25 mg L-carnitine/kg (8 replications per treatment) from 21 weeks of age. 2. Hens were inseminated with semen from Ross broiler breeder males. In a common facility, subsequent progeny hatchability and embryonic mortality at 25, 30, 32, and 38 weeks of breeder age were evaluated. 3. Subsequent egg component weights, incubational egg water loss, progeny embryo growth, and embryo, yolk sac and liver composition through 18 d of incubation at 27, 32, and 38 weeks of breeder age were evaluated. 4. Calculated additions of L-carnitine were in agreement with analysed contents of 3.5 and 31.1 mg free L-carnitine/kg of diet, respectively, and total L-carnitine concentrations increased by 48.6, 21.7, and 10.0% in 0-d yolk, 18-d yolk sac, and 18-d liver samples, respectively, due to the addition of dietary L-carnitine. 5. Supplemental L-carnitine resulted in increased (0.6%) relative 0-d egg yolk weight across weeks 27, 32, and 38, and reduced (0.38%) 18-d yolk sac palmitoleic acid concentration at week 27 without altering embryogenesis. 6. In conclusion, dietary L-carnitine (25 mg/kg of the diet) was deposited in the yolks of broiler breeder hens and was subsequently transferred to the embryonic liver via yolk sac absorption through 18 d of incubation. Furthermore, dietary L-carnitine supplementation increased ovarian follicle yolk deposition in 27-, 32-, and 38-week-old breeder hens, and influenced yolk sac fatty acid beta-oxidation in embryos from 27-week-old breeder hens causing yolk sac palmitoleic acid concentrations to be reduced by 18 d of incubation.

  19. OPTIMASI GEOMETRI TERAS REAKTOR DAN KOMPOSISI BAHAN BAKAR BERBENTUK BOLA PADA DESAIN HIGH TEMPERATURE FAST REACTOR (HTFR

    Directory of Open Access Journals (Sweden)

    Agustina Mega

    2015-04-01

    Full Text Available Telah dilakukan desain High Temperature Fast Reactor (HTFR tipe pebble dengan bahan bakar uranium plutonium nitrida berpendingin Pb-Bi. Parameter yang dianalisis adalah kritikalitas teras, koefisien reaktivitas temperatur bahan bakar, koefisien reaktivitas void pendingin dan kemampuan breeding reaktor. Perhitungan dilakukan dengan paket program SRAC2K3. Dari penelitian ini diharapkan diperoleh desain teras berumur lama dan memiliki fitur keselamatan melekat. Dari penelitian ini diperoleh desain reaktor dengan diameter 520 cm dan tinggi 480 cm. Bahan bakar berbentuk pebble dengan 63 % UN-37 % PuN pada zona core dan 95,5 % UN-4,5 % PuN pada zona blanket. Reaktor tidak kritis setelah kurang lebih 800 hari dan keff pada BoL 1,078223 dan keff setelah 800 hari adalah 0,986379. Dari penelitian ini diperoleh koefisien reaktivitas temperatur bahan bakar sebesar -2,190014E-05 pada saat BoL dan -1,390773E-05 setelah 800 hari serta koefisien reaktivitas void pendingin sebesar -2,160402E-04/% void pada saat BoL dan setelah 800 hari sebesar -2,942364E-03/% void. Reaktor merupakan jenis fast breeder ditandai dengan naiknya densitas plutonium 239. Kata kunci : desain, teras, HTFR, keselamatan, umur, koefisien reaktivitas.   Design of pebble bed type High Temperature Fast Reactor (HTFR with uranium plutonium nitride fuel and Pb-Bi cooled has been done. The parameters being analyzed were core criticality, fuel temperature coefficient, void coefficient and reactor breeding ability. Calculation was done by using SRAC2K3 computer code. This research is expected to obtaine the design with long life core and inherent safety features. This research obtained core design with a diameter of 520 cm and 480 cm core high. Shaped pebble fuel bed with the 63 % UN-37 % PUN on core zone and 95.5 % UN-4.5 % Pu on blanket zone and keff value is 1.078223 with approximately 800 day of core life. The fuel temperature coefficient is -2.190014E-05 at BOL and is 1.390773E-05 at EOL and

  20. Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Institute Regional des Materiaux Avances, Ispra (Italy); Casini, G. [Systems Engineering and Informatics Institute, JRC Ispra, Ispra (Vatican City State, Holy See) (Italy); Viola, A. [Department of Chemical Engineering, University of Cagliari, Cagliari (Italy)

    1995-03-01

    The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.).

  1. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  2. Axion Cooling of Neutron Stars

    CERN Document Server

    Sedrakian, Armen

    2015-01-01

    Cooling simulations of neutron stars and their comparison with the data from thermally emitting X-ray sources puts constraints on the properties of axions, and by extension of any light pseudo-scalar dark matter particles, whose existence has been postulated to solve the strong-CP problem of QCD. We incorporate the axion emission by pair-breaking and formation processes by $S$- and $P$-wave nucleonic condensates in a benchmark code for cooling simulations as well as provide fit formulae for the rates of these processes. Axion cooling of neutron stars has been simulated for 24 models covering the mass range 1 to 1.8 solar masses, featuring non-accreted iron and accreted light element envelopes, and a range of nucleon-axion coupling. The models are based on an equation state predicting conservative physics of superdense nuclear matter that does not allow for onset of fast cooling processes induced by phase transitions to non-nucleonic forms of matter or high proton concentration. The cooling tracks in the tempe...

  3. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  4. Formation of the research competence of future breeder-geneticists during the practical training

    Directory of Open Access Journals (Sweden)

    Наталія Павлівна Антіпова

    2016-03-01

    Full Text Available In the article were analyzed content, structure and special features of the practical training of the future breeder-geneticists. There was specified that for bachelor breeder-geneticists of the 6.090101 “Agronomy” training direction are provided the following practices: botany, tractors and cars, soil science, agricultural machines, fruit growing, vegetable growing, agriculture, agrochemistry, plant growing, entomology, phytopathology, technology of retreatment and preservation of plant growing production, selection and seed growing of agricultural crops; work practice on agronomy.In the article was grounded the structural-logic scheme of organization of practices of the future breeder-geneticists for forming their professional research competences. There were analyzed functions and methods of scientific-cognitive activity, forms, means and methods of the practical education. According to the types of practices there were elucidated and grounded the structure of practical training of the future breeder-geneticists – natural-technical and special-professional – educational and work practices.The natural-technical, special-professional and production-technological practices form in bachelor agronomists breeder-geneticists professional scientific-research competences: theoretical-analytical ones – geosphere, bio-geo-physical-chemical, meteorological-climatological, informative, fundamental and applied ecological, fundamental-scientific agrosphere; professional competences of professionally oriented and special practical training – agricultural, branch, techno-service, technological, selective, special-genetic-engineering ones; managerial ones

  5. SEROLOGICAL STATUS OF EGG DROP SYNDROME IN BREEDERS AND COMMERCIAL LAYERS IN MANSEHRA DISTRICT

    Directory of Open Access Journals (Sweden)

    S. T. BADAR1, M. SIDDIQUE, R. ALI1 AND M. H. RASOOL

    2006-01-01

    Full Text Available A serological survey was conducted to see the prevalence of antibodies to egg drop syndrome (EDS virus in breeders and commercial layers in Mansehra district. A total of 50 flocks including broiler breeder (30, layer breeder (10 and commercial layers (10 were surveyed from September, 2002 to March, 2003. Paired serum samples were collected at random from 20 birds in each flock at fortnight interval and screened through haemagglutination inhibition (HI test. In broiler breeders, out of 30 flocks, 12(40% were serologically positive for EDS. In layer breeders and commercial layers, 30 and 20% flocks were found to be seropositive, respectively. On the basis of age, maximum seroprevalence (42.85% was observed in birds between 21-40 weeks of age. The seroprevalence was 33.33% in 1-20 weeks, 26.66% in 41-60 weeks and 25.00 % in 61 weeks and above old birds. No specific clinical signs were observed in all three types of birds. A drop in egg production (22.555% was observed in seropositive flocks along with misshapen soft shelled, shell-less and non-pigmented eggs. Based on these findings, the farmers were advised to vaccinate their flocks against EDS between 14-16 weeks of age using killed adjuvanted vaccines.

  6. Progress of the stochastic cooling system of the Collector Ring

    CERN Document Server

    Dimopoulou, C; Bohm, R; Dolinskyy, O; Franzke, B; Hettrich, R; Maier, W; Menges, R; Nolden, F; Peschke, C; Petri, P; Steck, M; Thorndahl, L

    2013-01-01

    An overview of the recent achievements and ongoing developments for the stochastic cooling system of the Collector Ring is given. In focus are the hardware developments as well as the progress in predicting the system performance. The system operates in the frequency band 1-2 GHz, it has to provide fast 3D cooling of antiproton, rare isotope and stable heavy ion beams. The main challenges are (i) the cooling of antiprotons by means of cryogenic movable pick-up electrodes and (ii) the fast two-stage cooling (pre-cooling by the Palmer method, followed by the notch filter method) of the hot rare isotope beams (RIBs). Recently, a novel code for simulating the cooling process in the time domain has been developed at CERN. First results for the momentum cooling for heavy ions in the CR will be shown in comparison with results obtained in the frequency domain with the Fokker-Planck approach.

  7. 400 Area/Fast Flux Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The 400 Area at Hanford is home primarily to the Fast Flux Test Facility (FFTF), a DOE-owned, formerly operating, 400-megawatt (thermal) liquid-metal (sodium)-cooled...

  8. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  9. Measuring motivation for appetitive behaviour: food-restricted broiler breeder chickens cross a water barrier to forage in an area of wood shavings without food.

    Directory of Open Access Journals (Sweden)

    Laura M Dixon

    Full Text Available Broiler breeders (parents of meat chickens are selected for fast growth and become obese if fed ad libitum. To avoid this and maintain good health and reproductive ability, they are feed restricted to about 1/3 of what they would eat ad libitum. As a result, they experience chronic hunger and exhibit abnormal behaviour patterns that may indicate stress and frustration. One approach to measuring hunger is to observe how much birds will work, such as pecking a key, for access to more or different types of food. However, the sight, smell, and feedback from consumption of the feed reward changes the context and may artificially raise feeding motivation. To avoid this, we tested broiler breeders in an apparatus in which they could work for access to a wooden platform covered in wood shavings by crossing a water runway which increased in length and depth in 8 successive tests. In the wood shavings area, they could perform exploratory and foraging behaviour (the appetitive phase of feeding but were never rewarded with feed. Sixty birds were divided into three feed quantity treatments: commercial restriction (R, and twice (2R or three times (3R this amount. Overall, birds fed R worked harder to reach the wood shavings area (reached it in a larger number of tests than 2R and 3R birds (P2R>3R. This indicates that restricted-fed birds were hungry and willing to work for the opportunity to forage even though food was never provided, suggesting that their motivation to perform the appetitive component of feeding behaviour (foraging/food searching was sufficient to sustain their response. Thus food restriction in broiler breeders is a welfare concern. However these methods could be used to test alternative feeding regimes to attempt to find ways of alleviating hunger while still maintaining healthy growth and reproduction in these birds.

  10. Measuring motivation for appetitive behaviour: food-restricted broiler breeder chickens cross a water barrier to forage in an area of wood shavings without food.

    Science.gov (United States)

    Dixon, Laura M; Brocklehurst, Sarah; Sandilands, Vicky; Bateson, Melissa; Tolkamp, Bert J; D'Eath, Rick B

    2014-01-01

    Broiler breeders (parents of meat chickens) are selected for fast growth and become obese if fed ad libitum. To avoid this and maintain good health and reproductive ability, they are feed restricted to about 1/3 of what they would eat ad libitum. As a result, they experience chronic hunger and exhibit abnormal behaviour patterns that may indicate stress and frustration. One approach to measuring hunger is to observe how much birds will work, such as pecking a key, for access to more or different types of food. However, the sight, smell, and feedback from consumption of the feed reward changes the context and may artificially raise feeding motivation. To avoid this, we tested broiler breeders in an apparatus in which they could work for access to a wooden platform covered in wood shavings by crossing a water runway which increased in length and depth in 8 successive tests. In the wood shavings area, they could perform exploratory and foraging behaviour (the appetitive phase of feeding) but were never rewarded with feed. Sixty birds were divided into three feed quantity treatments: commercial restriction (R), and twice (2R) or three times (3R) this amount. Overall, birds fed R worked harder to reach the wood shavings area (reached it in a larger number of tests) than 2R and 3R birds (P2R>3R). This indicates that restricted-fed birds were hungry and willing to work for the opportunity to forage even though food was never provided, suggesting that their motivation to perform the appetitive component of feeding behaviour (foraging/food searching) was sufficient to sustain their response. Thus food restriction in broiler breeders is a welfare concern. However these methods could be used to test alternative feeding regimes to attempt to find ways of alleviating hunger while still maintaining healthy growth and reproduction in these birds.

  11. Optimization of electron-cyclotron-resonance charge-breeder ions : Final CRADA Report.

    Energy Technology Data Exchange (ETDEWEB)

    Pardo, R.; Physics; Far-Tech, Inc.

    2009-10-09

    Measurements of 1+ beam properties and associated performance of ECR Charge Breeder source determined by total efficiency measurement and charge state distributions from the ECR Charge Breeder. These results were communicated to Far-Tech personnel who used them to benchmark the newly developed programs that model ion capture and charge breeding in the ECR Charge Breeder Source. Providing the basic data described above and in the discussion below to Far-Tech allowed them to improve and refine their calculational tools for ECR ion sources. These new tools will be offered for sale to industry and will also provide important guidance to other research labs developing Charge Breeding ion sources for radioactive beam physics research.

  12. Unidentified Factors in Jojoba Meal Prevent Oviduct Development in Broiler Breeder Females.

    Science.gov (United States)

    Vermaut; Onagbesan; Bruggeman; Verhoeven; Berghman; Flo; Cokelaere; Decuypere

    1998-01-19

    Supplementation of feed with jojoba meal, as a means for autonomous feed restriction, was successful in depressing feed intake and controlling body weight of broiler breeder pullets to the extent recommended by the breeder company. However, these broiler breeders never produced eggs. At the level of ovary, normal follicle development and maturation did occur. A considerable number of ovulations occurred which were not followed by oviposition. After ovulation, the ova could not be captured by the oviduct, because of the small size of the oviduct, resulting in "internal laying". The virtual absence of oviduct development cannot be explained presently but it must be due to some yet unidentified factor(s) in jojoba meal which prevent(s) the normal development of the oviduct. These factors may be acting by abnormally increasing plasma progesterone or triiodothyronin levels and/or directly by themselves interfering with oviduct development. The nature of these factors requires further investigations.

  13. Be-Breeder - Learning: a new tool for teaching and learning plant breeding principles

    Directory of Open Access Journals (Sweden)

    Roberto Fritsche-Neto

    2016-09-01

    Full Text Available The Be-Breeder application is an on-line tool constructed through the R software for the purpose of assisting in some of the main genetic and statistical analyses related to the area of plant breeding. In addition, Be-Breeder provides a section called “Learning”, which in a simple click-point manner allows explanation of theories related to the effect of inbreeding, population structure, qualitative and quantitative traits, heterosis, population size, effect of selection, and composition of hybrids. Be-Breeder is available for network use on the website of the Allogamous Plant Breeding Laboratory (Laboratório de Melhoramento de Plantas Alógamas of ESALQ-USP through the link: http://www.genetica.esalq.usp.br/alogamas/R.html.

  14. Q Fever (Coxiella burnetii) Knowledge and Attitudes of Australian Cat Breeders and Their Husbandry Practices.

    Science.gov (United States)

    Shapiro, A J; Norris, J M; Bosward, K L; Heller, J

    2016-09-13

    A Q fever outbreak in a small animal veterinary hospital, associated with a cat caesarean section, initiated a cat seroprevalence study (n = 712) that found circulating antibodies to Coxiella burnetii was highest in cattery-confined breeding cats (9.3%). These findings stimulated interest about potential sources of C. burnetii infection for cats and humans associated with cats. Cat breeders are potentially a group at increased risk of C. burnetii infection, and this study sought to identify potential risk factors. A cross-sectional online survey was conducted targeting all domestic cat breeders registered with an affiliate member body in Australia in 2015. Responses from 177 cat breeders across Australia were analysed. Forty per cent of responding cat breeders had not heard of Q fever. Raw meat was fed as an integral constituent of the diet by 89% of respondents. Eighty per cent of respondents allowed queens access to the home for parturition, and assistance of queens and resuscitation of kittens at the time of birth were reported by 97% of respondents. Respondents who perceived some level of exposure to Q fever through their breeding activities were three times less likely to perform mouth-to-snout resuscitation (OR 0.3 95% CI 0.1-0.9; P = 0.034) than those who did not perceive a risk of exposure. Similarly, respondents who perceived Q fever as a risk through breeding activities were close to eight times more likely to use personal protective equipment during parturition (OR 7.7 95% CI 1.5-39.9; P = 0.015) than those who did not. Husbandry practices of cat breeders that may increase the risk of C. burnetii transmission require further targeted investigations to assess the contribution of these risk factors to the acquisition of disease. Concurrent education forums are recommended to inform Australian cat breeders of the aetiopathogenesis of Q fever.

  15. SALINOMYCIN TOXICOSIS IN BROILER BREEDERS AND TURKEYS: REPORT OF THE FIRST CASE

    Directory of Open Access Journals (Sweden)

    Konstantinos C. Koutoulis

    2013-01-01

    Full Text Available The objective of this study was to describe two clinical cases of toxicosis by accidental administration of salinomycin in feed in broiler breeders and turkeys. In the first clinical case, a 32000 broiler breeders flock, 30 weeks of age, suffered a severe mortality during the first week, a sharp decrease in egg production and a dramatic reduction in food consumption together with an increase of eating time. The pathological condition manifested itself mainly with leg paralysis and reluctance to move. In necropsy, signs of bile hyper-secretion, absence of feed in the digestive tract, liver degeneration, egg peritonitis, enlarged kidneys and spleen atrophy were found. Serological examinations for IBV, IBD, NDV, Mg, Ms, Ort, TRT, AE and CAA were negative. The broiler breeders flock was finally euthanized due to an irreversible condition, slow recovery and poor production indicators. In the second clinical case, a flock of 27000 turkeys fatteners, aged from 20 to 35 days, showed an increased daily mortality up to 10 and a 88% total death loss in 2 weeks period. The turkeys showed great growth retardation, reluctance to move and eat, ataxia and absence of gross lesions, except atrophy of the spleen and pica. Serological examinations for NDV and AI were negative. After removal and change of feed, the remaining flock recovered normally. These two clinical cases were initially associated with incorrect diet and, after obtaining detailed clinical history, laboratory tests and food analysis, the by error administration of salinomycin in feed was confirmed. The final concentrations of salinomycin in feed analysis reached 64.6-124 mg kg-1 for the broiler breeders and 298-944 mg kg-1 for turkeys. While salinomycin toxicosis has been reported worldwide in the literature in a number of animal species, including turkey breeders and fatteners with similar symptoms, in broiler breeders has never been reported on the field.

  16. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  17. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  18. BN800: The advance sodium cooled fast reator plant based on dose fuel sycle%BN800:定位于闭式燃料循环的先进钠冷快堆核电站

    Institute of Scientific and Technical Information of China (English)

    吴兴曼

    2011-01-01

    Abstract: As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc.%作为实际上快堆技术最先进的国家之一,俄罗斯始终站在快堆技术发展的前沿.在成功运行了电功率为600MW的BN600快堆核电站近30年,以及在其基础上改进并完成数次设计优化数十年后,终于决定建设别洛雅尔斯基核电站4号机组(BN800快堆电站).BN800快堆核电站被认为是世界上正在付诸工程的最为先进的快堆核电机组.我国的快堆技术发展

  19. Leptin receptor in the chicken ovary: potential involvement in ovarian dysfunction of ad libitum-fed broiler breeder hens

    Directory of Open Access Journals (Sweden)

    Collin Anne

    2004-10-01

    Full Text Available Abstract In hens, the ovarian follicles committed to ovulation are arranged in an ordered follicular hierarchy. In standard broiler breeders hens genetically selected for high growth rate the reproductive function is clearly dysfunctional. Feed restriction is needed during reproductive development to limit the formation of excessive numbers of ovarian yellow follicles arranged in multiple hierarchies. To determine whether leptin is involved in the nutritional and reproductive interactions controlling follicular hierarchy in hens, blood leptin levels and ovarian expression of the leptin receptor mRNA were determined during follicle maturation in three chicken lines; a slow growing broiler "Label" genotype without reproductive dysfunction, a fast growing "Standard" genotype fed ad libitum or restricted and a fast growing "Experimental" line with intermediate reproductive performance levels. Whereas expression of the leptin receptor mRNA did not change in the theca, it clearly decreased with follicular differentiation in the granulosa of slow growing hens. In fast growing standard hens fed ad libitum and presenting significant reproductive dysfunction, the decrease was disrupted and dramatic up-regulation of granulosa cell expression of the leptin receptor was observed. On the other hand, feed restriction decreased the overall level of expression of the leptin receptor mRNA and restored the decrease with follicular growth. The level of expression of the leptin receptor probably modulates the action of leptin on follicular differentiation. Since blood leptin and other metabolic factors were not affected by the genotype or by nutritional state, the factors involved in the regulation of leptin receptor gene expression remain to be determined. This study demonstrates the involvement of leptin in the nutritional control of reproduction in birds. Leptin action on the ovary probably controls follicular hierarchy through the regulation of steroidogenesis.

  20. Research and Implementation of Sodium-Cooled Fast Reactor Engineering SOA Integration Platform Based Upon Information System%基于SOA的钠冷快堆工程信息系统集成平台的研究和实现

    Institute of Scientific and Technical Information of China (English)

    李健

    2014-01-01

    This paper analyzes the implementation of integrated technology and its limitations exist in the current construction of nuclear power construction management system, and for the sodium cooled fast reactor construction management features, proposed to build SOA-based cold sodium fast reactor engineering information system integration platform (hereinafter referred to as "integration Platform") design ideas. System integration platform for business functions, and system integration theory, the concepts of SOA, analysis and design in-depth study on the basis of the design of the overall structure of the sodium cooled fast reactor engineering information system integration platform, and discusses in detail the integrated platform-specific implementation process.%本文主要分析了当前核电建造施工管理系统集成的实现技术及其存在的局限性,并针对钠冷快堆的施工管理特点,提出了构建基于SOA的钠冷快堆工程信息系统集成平台(以下简称“集成平台”)的设计思想。对集成平台的业务功能体系、系统集成的理论、SOA的相关概念、分析与设计进行了深入研究,在此基础上设计了钠冷快堆工程信息系统集成平台的总体架构,并详细论述了集成平台的具体实现过程。

  1. Effect of scattered feeding and feeding twice a day during rearing on indicators of hunger and frustration in broiler breeders

    NARCIS (Netherlands)

    Jong, de I.C.; Wolthuis-Fillerup, M.; Blokhuis, H.J.

    2005-01-01

    Broiler breeders are routinely fed restricted during rearing which has a major negative effect on their welfare. They suffer from hunger and frustration from thwarting of feeding. The aim of this experiment was therefore to study if broiler breeder welfare can be improved by changes in the feeding s

  2. The Effect of Low-Density Diets on Broiler Breeder Development and Nutrient Digestibility During the Rearing Period

    NARCIS (Netherlands)

    Enting, H.; Veldman, A.; Verstegen, M.W.A.; Aar, van de P.J.

    2007-01-01

    Low-density diets might help to reduce hunger feeling in restricted-fed broiler breeders. Effects of low-density diets on nutrient digestibility and bird development were studied in Cobb 500 broiler breeder hens from 4 wk of age until the onset of the lay (wk 26). The experiment included 4 treatment

  3. Breeding business : the future of plant breeding in the light of developments in patent rights and plant breeder's rights

    NARCIS (Netherlands)

    Louwaars, N.P.; Dons, J.J.M.; Overwalle, van G.; Raven, H.; Arundel, A.; Eaton, D.; Nelis, A.

    2009-01-01

    Plant breeding serves an important public interest. Two intellectual property (IP) systems are relevant for the protection of innovations in this sector: plant breeder's rights and patent rights. Some exemptions play an important role in plant breeding, such as the 'breeder's exemption', which is

  4. Breeding business : the future of plant breeding in the light of developments in patent rights and plant breeder's rights

    NARCIS (Netherlands)

    Louwaars, N.P.; Dons, J.J.M.; Overwalle, van G.; Raven, H.; Arundel, A.; Eaton, D.; Nelis, A.

    2009-01-01

    Plant breeding serves an important public interest. Two intellectual property (IP) systems are relevant for the protection of innovations in this sector: plant breeder's rights and patent rights. Some exemptions play an important role in plant breeding, such as the 'breeder's exemption', which is un

  5. Energy utilization and heat production of embryos from eggs originating from young and old broiler breeder flocks

    NARCIS (Netherlands)

    Nangsuay, A.; Meijerhof, R.; Ruangpanit, Y.; Kemp, B.; Brand, van den H.

    2013-01-01

    Two experiments were conducted to study the interaction between breeder age and egg size on the energy utilization (experiment 1) and heat production (experiment 2) of broiler embryos. In experiment 1, a total of 4,800 Ross-308 hatching eggs from 2 breeder ages (29 and 53 wk of age, or young and old

  6. Effects of incubational humidity and hen age on embryo composition in broiler hatching eggs from young breeders.

    Science.gov (United States)

    Peebles, E D; Burnham, M R; Gardner, C W; Brake, J; Bruzual, J J; Gerard, P D

    2001-09-01

    There is a paucity of information regarding the influence of incubational humidity on the characteristics of broiler embryos from young hens that often produce chicks of marginal quality. Therefore, the effects of broiler breeder age (26, 28, and 30 wk) in young broiler breeders and incubator humidity (43, 53, and 63% RH) on embryonic growth between 16 and 21 d of incubation, hatching chick weight, and embryo and chick body compositions were evaluated. Hatching broiler chick BW at 21 d was lower for 26-wk-old breeders compared to those at 28 and 30 wk, and the lowest RH decreased relative wet embryo weight. Embryonic growth, on wet and DM bases, was increased in eggs from the youngest hens. Embryo moisture content (EM) was lower at 16, 17, and 19 d of incubation in eggs from breeders at 26 wk when compared to those at 28 wk of age. The effects of RH on embryo crude protein and fat composition were not consistent and varied with breeder age and length of incubation. It was demonstrated that broiler breeder age affects embryogenesis and hatching chick BW. Furthermore, EM was influenced by breeder age, and changes in incubational RH between 43 and 63% affected embryogenesis without having associated effects on EM or consistent effects on embryo crude fat and protein content. Nevertheless, a depression in embryogenesis with a reduction in incubational RH to 43% may accentuate poor posthatch performance of chicks from young breeders.

  7. Recent Advance in Thermoelectric Devices for Electronics Cooling

    Science.gov (United States)

    Wang, Peng

    Thermal management of on-chip hot spot, with a heat flux of around 1000 W/cm2, has become one of the major challenges in the development of next-generation microprocessors. Solid state thermoelectric cooler (TEC) offers great promise for hot spot thermal management because of their compact structure, fast response, high reliability, localized cooling, and high flux removal capability. To date TEC has received great attentions in electronics cooling community as one of the potential hot spot cooling solutions. In this paper, recent development and application of hot spot cooling strategies based on micro thermoelectric technologies will be reviewed and discussed, three hot spot cooling concepts, including thinfilm thermoelectric cooling, mini-contact cooling, and semiconductor selfcooling in silicon substrate and germanium substrate will be discussed. The advantages and disadvantages of these on-chip cooling solutions for high flux hot spots will be evaluated.

  8. Hybrid radiator cooling system

    Science.gov (United States)

    France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

    2016-03-15

    A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

  9. Bone mineral density of tibae and femura of broiler breeders: growth, development and production

    Directory of Open Access Journals (Sweden)

    ICL Almeida Paz

    2006-06-01

    Full Text Available The aim of this study was to follow-up the physiological variations in the development of the bone tissue, associating them with the egg production curve. This study was carried out in the facilities of the Faculdade de Medicina Veterinária e Zootecnia of the UNESP, Botucatu, Brazil. Twenty-three families of Ross broiler breeders were used, each family consisting of 13 females and 1 male, distributed in 23 pens of 5.0m² each. The management was that recommended by the genetic company manual (Agroceres Ross, 2003, with daily feeding until 6th week of age; and birds were fed according to a 5:2 schedule (5 days fed, 2 days of fasting between 7 and 17 weeks of age, returning to daily feeding starting at 18 weeks of age. Birds did not receive afternoon calcium supplementation. On the fourth week of rearing, 84 females were removed for bone analyses of the right tibia and femur, using optical densitometry in radiographic images technique. These analyses were sequentially carried out in 4, 8, 12, 15, 20, 24, 30, 35, 42, 47, and 52 week-old birds. The egg production curve of the birds was followed-up and associated to bone mineral density results. For bone mineral density evaluation (BMD birds were divided by weight categories as light, intermediate, or heavy within each data age. BMD values of the tibias were not influenced by weight range, but by the age at collection. On the other hand, interactions were found among femur BMD values and weight and age categories. There was no correlation between eggshell quality and femur BMD. A negative correlation (-0.15 was observed between tibia BMD and eggshell percentage. It was possible to conclude that the egg production has little influence on bone mineral density of the birds probably because there was no need of bone mineral mobilization during the production period, since the observed egg production was below that observed under commercial conditions.

  10. Current progress in laser cooling of antihydrogen

    CERN Document Server

    Luschevskaya, E V

    2014-01-01

    We talk about laser cooling methods of $\\bar{H} (H)$ and experimental achievements in this area. The Lyman-$\\alpha$ transition $1S\\rightarrow 2P$ is the most suitable one for this purpose due to small lifetime of $2P$ state and insignificant ionization losses at this wavelength. However the pulsed and continuous laser sources of Lyman-$\\alpha$ wavelength do not have enough power for fast cooling. The absence of powefull sources of Lyman-$\\alpha$ irradiation is a technical problem associated with the complexity of four-wave mixing scheme used for irradiation generation. Another problem of $\\bar{H} (H)$ laser cooling is large recoil energy which prevents cooling to submiliKelvin range by simple methods. The alternative way could be in use of other spectral transitions in (anti)hydrogen, new laser cooling methods or even other cooling schemes such as boofer or sympathetic cooling. In this paper we also discuss the applicability of sympathetic cooling for the antihydrogen case. The exploration of antihydrogen and...

  11. Image Inpainting by Cooling and Heating

    DEFF Research Database (Denmark)

    Gustafsson, David Karl John; Pedersen, Kim Steenstrup; Nielsen, Mads

    2007-01-01

    We discuss a method suitable for inpainting both large scale geometric structures and stochastic texture components. We use the well-known FRAME model for inpainting. We introduce a temperature term in the learnt FRAME Gibbs distribution. By using a fast cooling scheme a MAP-like solution is found...

  12. Effect of low-density diets on broiler breeder and offspring performance

    NARCIS (Netherlands)

    Enting, H.

    2005-01-01

    Restricted feeding of broiler breeders is required to obtain good reproductive performance. Current practical feed restriction levels can result in hunger feeling and chronic stress, particularly during the reari