Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Peterson, Per; Greenspan, Ehud

2015-01-01

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Fuel and helium confinement in fusion reactors

International Nuclear Information System (INIS)

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

1993-01-01

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

Design of a spherical fuel element for a gas-cooled fast reactor

International Nuclear Information System (INIS)

Van Rooijen, W.F.G.; Kloosterman, J.L.; Van Dam, H.; Van der Hagen, T.H.J.J.

2004-01-01

A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k eff with burn-up. Nitride fuel ( 15 N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

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

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

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

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

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

The modular high-temperature gas-cooled reactor - a new production reactor

International Nuclear Information System (INIS)

Nulton, J.D.

1990-01-01

One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

International Nuclear Information System (INIS)

Monteith, H.C.

1978-10-01

The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-01-01

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

Gas Cooled Fast Reactors: Recent advances and prospects

International Nuclear Information System (INIS)

Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

2013-01-01

Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

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

International Nuclear Information System (INIS)

Wang Jie; Gu Yihua

2004-01-01

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

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).

The evolution of US helium-cooled blankets

International Nuclear Information System (INIS)

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

1991-01-01

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

Fuel arrangement for high temperature gas cooled reactor

International Nuclear Information System (INIS)

Tobin, J.M.

1978-01-01

Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

Cooling nuclear reactor fuel

International Nuclear Information System (INIS)

Porter, W.H.L.

1975-01-01

Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

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

Science.gov (United States)

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

1975-01-01

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

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

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

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

High power density reactors based on direct cooled particle beds

Science.gov (United States)

Powell, J. R.; Horn, F. L.

Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-11-15

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Helium-cooled nuclear reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

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

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

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

Development of gas-cooled fast reactor and its thermo-hydraulics

International Nuclear Information System (INIS)

Kawamura, Hiroshi

1977-10-01

Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide

International Nuclear Information System (INIS)

Thompson, W.I.

1980-10-01

CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels

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.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

International Nuclear Information System (INIS)

Shropshire, D.E.; Herring, J.S.

2004-01-01

The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

International Nuclear Information System (INIS)

Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

2012-01-01

After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

1975-06-01

The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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

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

International Nuclear Information System (INIS)

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

1998-03-01

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

Development of components for the gas-cooled fast breeder reactor program

International Nuclear Information System (INIS)

Dee, J.B.; Macken, T.

1977-01-01

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-11-01

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

Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

International Nuclear Information System (INIS)

Rooijen, W.F.G. van

2006-01-01

The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

Gas turbine modular helium reactor in cogeneration

International Nuclear Information System (INIS)

Leon de los Santos, G.

2009-10-01

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

An alternative solution for heavy liquid metal cooled reactors fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

CEA programme on gas cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

2002-01-01

partly builds on the past experience of the CEA on gas cooled reactors (UNGG, HTR) and on the current effort to revive and update High Temperature Reactor technologies to support the development of modular helium cooled reactors (∼300 MWe) by Framatome-ANP and international partners. In this context, the CEA decided to focus prospective R and D work on the development of a consistent set of gas cooled nuclear systems ranging from medium term reactor projects for electricity generation and other applications (robust and secure export model, process heat, hydrogen production at very high temperature, plutonium burning) to a longer term vision of sustainable nuclear systems using fast neutrons with a closed and integrated fuel cycle. This range of gas cooled nuclear systems covers a wide variety of high temperature applications as well as a broad range of fuel cycles, including synergistic fuel cycles with light water reactors (i.e. burning plutonium and possibly also minor actinides from PWR spent fuels). A specific research and development programme is being currently implemented to support the development of this consistent set of gas cooled systems. The major emphasis is put on fuel particles re-fabrication and possible adaptations to fast neutrons, on high temperature materials, high temperature systems technology, and compact spent fuel processing and re-fabrication processes. This programme anticipates the construction of large experimental facilities in the next decade, such as an He integral test loop (2007), a technology testing reactor and a lab scaled integrated fuel cycle (2012). A substantial effort is also invested in the validation of computational tools and procedures for feasibility and performance studies. Strong connections with fundamental research (nuclear physics, materials science, nuclear chemistry) are essential to improve the modelling capability and to achieve effective breakthroughs for the development of high temperature and high irradiation

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

International Nuclear Information System (INIS)

Han, Kyu Hyun; Chang, Soon Heung

2005-01-01

More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

Medium-size high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Peinado, C.O.; Koutz, S.L.

1980-08-01

This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-07-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

International Nuclear Information System (INIS)

1985-01-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

Thermal and flow design of helium-cooled reactors

International Nuclear Information System (INIS)

Melese, G.; Katz, R.

1984-01-01

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

Evaluation of helium cooling for fusion divertors

International Nuclear Information System (INIS)

Baxi, C.B.

1993-09-01

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

Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

International Nuclear Information System (INIS)

Silva, Alice Cunha da; Su, Jian

2013-01-01

The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

High power density reactors based on direct cooled particle beds

International Nuclear Information System (INIS)

Powell, J.R.; Horn, F.L.

1985-01-01

Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

International Nuclear Information System (INIS)

Silady, F.A.; Millunzi, A.C.

1989-08-01

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Fuel performance and fission product behaviour in gas cooled reactors

International Nuclear Information System (INIS)

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

Use of thorium for high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

2017-07-01

The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

2014-08-15

The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

Directory of Open Access Journals (Sweden)

Igor Shamanin

2015-01-01

Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

Fuel element replacement and cooling water activity at the musashi reactor

International Nuclear Information System (INIS)

Nozaki, Tetsuya; Honda, Teruyuki; Horiuchi, Norikazu; Aizawa, Otohiko; Sato, Tadashi

1989-01-01

The Musashi Institute of Technology Research Reactor (TRIGA 11, 100 kW) has been operated without serious problems since 1963. However, because there is no more spare fuel element, it was necessary to decide how to solve the problem. In the end, it was decided to obtain many stainless steel-clad fuel elements and operate with those fuel elements only, under the auspices of the Ministry of Education, Science and Culture. The bulk shielding experimental pool was remodeled as the storage for spent fuel elements, where the neutrons from the thermalizing column were shielded with cadmium and boron polyethylene plates. The equipment for transferring spent fuel elements was built and temporarily set up between the core tank and the new storage. These works were started in 1983, and finished in 1985. After the reactor was restarted, the count rate of the conventional cooling water monitor which was set in the cooling system using a GM counter drastically decreased. The spent fuel storage, the equipment and the works for fuel transfer, and the radioactivity of cooling water are reported. (K.I.)

Fuel performance and fission product behaviour in gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

Science.gov (United States)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

Directory of Open Access Journals (Sweden)

Grodzki Marcin

2017-12-01

Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Thermal insulation of the high-temperature helium-cooled reactors

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

Monado, F.; Permana, S.

2013-01-01

Full-text: 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. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto

2006-04-15

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

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

Tsuji, H.; Kondo, T.

1984-01-01

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

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

International Nuclear Information System (INIS)

Li Wenlong; Xie Heng

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

High-temperature gas-cooled reactors and process heat

International Nuclear Information System (INIS)

Kasten, P.R.

1980-01-01

High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

International Nuclear Information System (INIS)

Kulikov, I.S.; Tverkovkin, B.E.

1984-01-01

The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

Deep-Burn Modular Helium Reactor Fuel Development Plan

Energy Technology Data Exchange (ETDEWEB)

McEachern, D

2002-12-02

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

Thermal-hydraulics of actinide burner reactors

International Nuclear Information System (INIS)

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

1989-07-01

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

Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

International Nuclear Information System (INIS)

Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

2009-01-01

Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

International Nuclear Information System (INIS)

Li Junfeng

2016-01-01

High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

PG-100 helium loop in the MR reactor

International Nuclear Information System (INIS)

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

1983-01-01

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

Study on thermodynamic cycle of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu Xinhe; Yang Xiaoyong; Wang Jie

2017-01-01

The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Overview of environmental control aspects for the gas-cooled fast reactor

International Nuclear Information System (INIS)

Nolan, A.M.

1981-05-01

Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

Assessment of the dry process fuel sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

Assessment of the dry process fuel sodium-cooled fast reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

Cooling with Superfluid Helium

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-10-15

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

The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool; Analiza hladjenja ozracenog goriva u bazenu za odlaganje reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Vrhovac, M; Afgan, N; Spasojevic, D; Jovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1985-07-01

According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

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

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

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

CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

2005-01-01

High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

2016-07-15

Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

International Nuclear Information System (INIS)

Reyes-Ramirez, Ricardo; Martin-del-Campo, Cecilia; Francois, Juan-Luis; Brun, Emeric; Dumonteil, Eric; Malvagi, Fausto

2010-01-01

The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations. The main objective of this work is to compare the fuel depletion results obtained with MCNPX-CINDER90 code and the new TRIPOLI-4-Depletion code (developed by the Commissariat a l'Energie Atomique) of a fuel design concept for the Gas-cooled Fast Reactor. Calculations were made for an equivalent homogeneous model of fuel rods in a hexagonal mesh assembly. Total reflection conditions were applied on the six lateral faces and the two axial faces of the assembly. The materials used in the fuel assembly are: carbide of uranium and plutonium as fuel, silicon carbide as cladding, and helium gas as coolant. JEFF libraries of effective cross sections were used in both codes. Two methods of burnup step calculations were performed with TRIPOLI-4-D, the Euler and the CSADA, and their results were compared with the MCNPX-CINDER90 CSADA method. A period of 300 days of irradiation time was considered, which was divided into 12 steps. Results of the infinite multiplication factor as function of the irradiation time, and the evolution of the isotope concentrations for a selected group of nuclides were compared. The main conclusion is that very similar results were obtained for the three types of depletion calculations which were compared: (1) MCNPX-C90 CSADA; (2) TRIPOLI-4-D CSADA, and (3) TRIPOLI-4-D EULER. The best calculation time was obtained with the TRIPOLI-4-D EULER method, which needed approximately half the time than the other two. In summary, it is sufficiently good to use

Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

International Nuclear Information System (INIS)

Forsberg, C.

2015-01-01

The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

Material development for gas-cooled high temperature reactors for the production of nuclear process heat

International Nuclear Information System (INIS)

Nickel, H.

1977-04-01

In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Fuel assembly for gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Yellowlees, J.M.

1976-01-01

A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

International Nuclear Information System (INIS)

Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

2001-01-01

Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

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

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

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

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

International Nuclear Information System (INIS)

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

1978-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-11

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

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

Fuel element replacement and cooling water radioactivity at the Musashi reactor

International Nuclear Information System (INIS)

Nozaki, T.; Honda, T.; Horiuchi, N.; Aizawa, O.; Sato, T.

1988-01-01

The Musashi reactor (TRIGA-II, 100kW) has been operated without any serious troubles since 1963. In 1985 the old Al-cladded fuel elements were replaced with new stainless cladded ones in order to insure a long and safe operation. By using a semi-automatic equipment the old fuel elements have been transferred into the bulk-shielding experimental pool, which was remodelled for the spent-fuel storage. In order to reduce the exposure during the transfer work, the old fuel elements were cooled in the core tank for 3 months. After the replacement, the radioactivities in the cooling water have been drastically changed. The activity of Na-24 decreased about one decade, and the activities of Cr-51, Mn-54, Mn-56, Co-58 and Co-60 increased about two decades. At this conference we will report on the following points: (1) semi-automatic equipment for the transportation of the Al-cladded spent fuel, (2) structure of spent-fuel storage pool, and (3) radioactivity change in the cooling water. (author)

Development of high temperature gas cooled reactor in China

Energy Technology Data Exchange (ETDEWEB)

Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

2018-02-15

High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-07-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

International Nuclear Information System (INIS)

1988-01-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

USE OF THE MODULAR HELIUM REACTOR FOR HYDROGEN PRODUCTION

International Nuclear Information System (INIS)

SCHULTZ, K.R.

2003-01-01

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

Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

International Nuclear Information System (INIS)

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

2011-01-01

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 account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

Mirror hybrid reactor blanket and power conversion system conceptual design

International Nuclear Information System (INIS)

Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

1976-01-01

The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

High Temperature Gas Cooled Reactor Fuels and Materials

International Nuclear Information System (INIS)

2010-03-01

At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

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

International Nuclear Information System (INIS)

He, Qingyun; Feng, Jingchao; Chen, Hongli

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-02-15

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

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

International Nuclear Information System (INIS)

Huang, Jie; Ding, Ming

2017-01-01

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

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

International Nuclear Information System (INIS)

Halvers, L.J.

1977-03-01

An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

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

International Nuclear Information System (INIS)

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

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.

Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

International Nuclear Information System (INIS)

Casino, William A. Jr.

2006-01-01

Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

2013-03-01

'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Takase, Kazuyuki

1994-11-01

The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

Supercritical water-cooled reactor fuel management and economic comparison and analysis

International Nuclear Information System (INIS)

Cai Guangming; Ruan Liangcheng; Liu Xuechun

2014-01-01

The supercritical water-cooled reactor (SCWR) is expected to have an excellent fuel economical efficiency because of its high thermal efficiency. This article compares CSR1OOO with the current mainstream PWR and ABWR on the aspect of the economical efficiency of fuel management, and finally makes an unexpected conclusion that the SCWR has worse fuel economy than others. And it remains to be deliberated whether the SCWR will be the fourth generation of nuclear system. (authors)

Neutronic of heterogenous gas cooled reactors

International Nuclear Information System (INIS)

Maturana, Roberto Hernan

2008-01-01

At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

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

Test results from a helium gas-cooled porous metal heat exchanger

International Nuclear Information System (INIS)

North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

1996-01-01

A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

Fuel rod for liquid metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Vinz, P.

1976-01-01

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

ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

International Nuclear Information System (INIS)

Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

2008-01-01

In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

Science.gov (United States)

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

International Nuclear Information System (INIS)

2010-10-01

The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

International Nuclear Information System (INIS)

Mouri, Tomoaki; Kunitomi, Kazuhiko

2008-01-01

A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

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

OpenAIRE

Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

2016-01-01

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

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-12-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-06-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

Interdiffusion of krypton and xenon in high-pressure helium

International Nuclear Information System (INIS)

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

1980-01-01

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

Definition of breeding gain for the closed fuel cycle and application to a gas cooled fast reactor

International Nuclear Information System (INIS)

Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

2006-01-01

In this paper a definition is given for the Breeding Gain (BG) of a nuclear reactor, taking into account compositional changes of the fuel during irradiation, cool down and reprocessing. A definition is given for the reactivity weights required to calculate BG. To calculate the effects of changes in the initial fuel composition on BG, first order nuclide perturbation theory is used. The theory is applied to the fuel cycle of GFR600, a 600 MWth Generation IV Gas Cooled Fast Reactor. This reactor should have a closed fuel cycle, with a BG equal to zero, breeding just enough new fuel during irradiation to allow refueling by only adding fertile material. All Heavy Metal is recycled in the closed fuel cycle. The result is that a closed fuel cycle is possible if the reprocessing has low losses ( 238 U, 15% Pu, and low amounts of the Minor Actinides. (authors)

A combined gas cooled nuclear reactor and fuel cell cycle

Science.gov (United States)

Palmer, David J.

Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

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)

Comparison of fuel assemblies in lead cooled fast reactors

International Nuclear Information System (INIS)

Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G.

2016-09-01

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)

The Dragon reactor experiment

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

Inspection of fuel elements in the cooling pond of a research reactor

International Nuclear Information System (INIS)

Pavlov, S.V.; Mestnikov, A.V.

1992-01-01

Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh

2004-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

Science.gov (United States)

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

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.

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

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

Full Text Available 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.

Prestressed concrete vessels suitable for helium high temperature reactors

International Nuclear Information System (INIS)

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

1967-02-01

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

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

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

The continuous fuel cycle model and the gas cooled fast reactor

International Nuclear Information System (INIS)

Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

2011-01-01

The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

R + D work on gas-cooled breeder development

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

1978-01-01

The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Design and analysis of helium Brayton power cycles for HiPER reactor

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Design and analysis of helium Brayton power cycles for HiPER reactor

International Nuclear Information System (INIS)

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

2013-01-01

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

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

International Nuclear Information System (INIS)

Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

2012-01-01

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

International Nuclear Information System (INIS)

K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

2005-01-01

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 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 GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw; Venneri, Francesco

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Michael A. Pope

2011-10-01

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

Modeling and Application of Pneumatic Conveying for Spherical Fuel Element in Pebble-Bed Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

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

2014-01-01

The fuel handling system is an important system for on-load refueling in pebble-bed modular high-temperature gas-cooled reactor. A dynamic model of pneumatic conveying for spherical fuel element in fuel handling system was established to describe the pneumatically conveying process. The motion characteristics of fuel elements in pipeline and the effect of fuel elements on gas velocity were studied using the model. The results show that the theoretical analyses are consistent with the experimental. The research has been used in developing full scope simulator for pebble-bed modular high-temperature gas-cooled reactor, also provides references for the design and optimization of the fuel handling system. (author)

A comparative study on recycling spent fuels in gas-cooled fast reactors

International Nuclear Information System (INIS)

Choi, Hangbok; Baxter, Alan

2010-01-01

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

International Nuclear Information System (INIS)

Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

1976-05-01

For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie

1998-01-01

Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

International Nuclear Information System (INIS)

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

1981-11-01

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

International Nuclear Information System (INIS)

Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

1999-01-01

Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study

Overview of gas cooled reactors' applications with CATHARE

International Nuclear Information System (INIS)

Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

2005-01-01

Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

International Nuclear Information System (INIS)

Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

2004-01-01

Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

Study on the nitride fuel fabrication for FBR cycle (1)

International Nuclear Information System (INIS)

Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

2002-07-01

In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

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

International Nuclear Information System (INIS)

1979-04-01

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

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

1975-01-01

A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

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

International Nuclear Information System (INIS)

Cai Jianhua; Yang Xiaoyong; Wang Jie; Yu Suyuan

2008-01-01

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

CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket surrounding cell

International Nuclear Information System (INIS)

Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Shen, Feng; Yuan, Luzheng

2005-01-01

The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket around cell

International Nuclear Information System (INIS)

Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Cheng, Liang; Shen, Feng; Yuan, Luzheng

2005-01-01

The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

Design considerations for economically competitive sodium cooled fast reactors

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

2009-01-01

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

1982-01-01

Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-09-01

Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

Remote helium leak test of the DUPIC fuel rod

International Nuclear Information System (INIS)

Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

1998-01-01

DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

International Nuclear Information System (INIS)

Ganguly, Chaitanyamoy

2015-01-01

Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

Optimization of the fuel assembly for the Canadian Supercritical Water-cooled Reactor (SCWR)

Energy Technology Data Exchange (ETDEWEB)

French, C.; Bonin, H.; Chan, P., E-mail: Corey.French@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

A parametric optimization of the Canadian Supercritical Water-cooled Reactor (SCWR) lattice geometry and fresh fuel content is performed in this work. With the potential to improve core physics and performance, significant gains to operating and safety margins could be achieved through slight progressions. The fuel performance codes WIMS-AECL and SERPENT are used to calculate performance factors, and use them as inputs to an optimization algorithm. (author)

The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

International Nuclear Information System (INIS)

Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

1988-01-01

High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

The gas-cooled high temperature reactor: perspectives, problems and programmes

International Nuclear Information System (INIS)

Beckurts, K.H.; Engelmann, P.; Erb, D.E.

1977-01-01

For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

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

International Nuclear Information System (INIS)

Weaver, K.D.

2005-01-01

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)

Quantities of actinides in nuclear reactor fuel cycles

International Nuclear Information System (INIS)

Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)

Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1986-01-01

Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

Forced two phase helium cooling of large superconducting magnets

International Nuclear Information System (INIS)

Green, M.A.; Burns, W.A.; Taylor, J.D.

1979-08-01

A major problem shared by all large superconducting magnets is the cryogenic cooling system. Most large magnets are cooled by some variation of the helium bath. Helium bath cooling becomes more and more troublesome as the size of the magnet grows and as geometric constraints come into play. An alternative approach to cooling large magnet systems is the forced flow, two phase helium system. The advantages of two phase cooling in many magnet systems are shown. The design of a two phase helium system, with its control dewar, is presented. The paper discusses pressure drop of a two phase system, stability of a two phase system and the method of cool down of a two phase system. The results of experimental measurements at LBL are discussed. Included are the results of cool down and operation of superconducting solenoids

Optimization of TRU burnup in modular helium reactor

International Nuclear Information System (INIS)

Yonghee, Kim; Venneri, F.

2007-01-01

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

Advanced gas cooled nuclear reactor materials evaluation and development program

International Nuclear Information System (INIS)

1977-01-01

Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

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

International Nuclear Information System (INIS)

Forsberg, C.W.

2002-01-01

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

Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

International Nuclear Information System (INIS)

Skoda, R.; Rataj, J.; Uhera, J.

2008-01-01

The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

Energy Technology Data Exchange (ETDEWEB)

Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

1982-01-01

This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

International Nuclear Information System (INIS)

Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

1982-01-01

This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

Implications of alpha-decay for long term storage of advanced heavy water reactor fuels

International Nuclear Information System (INIS)

Pencer, J.; McDonald, M.H.; Roubtsov, D.; Edwards, G.W.R.

2017-01-01

Highlights: •Alpha decays versus storage time are calculated for examples of advanced heavy water reactor fuels. •Estimates are made for fuel swelling and helium bubble formation as a function of time. •These predictions are compared to predictions for natural uranium fuel. •Higher rates of damage are predicted for advanced heavy water reactor fuels than natural uranium. -- Abstract: The decay of actinides such as 238 Pu, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed.

Numerical modeling and validation of helium jet impingement cooling of high heat flux divertor components

International Nuclear Information System (INIS)

Koncar, Bostjan; Simonovski, Igor; Norajitra, Prachai

2009-01-01

Numerical analyses of jet impingement cooling presented in this paper were performed as a part of helium-cooled divertor studies for post-ITER generation of fusion reactors. The cooling ability of divertor cooled by multiple helium jets was analysed. Thermal-hydraulic characteristics and temperature distributions in the solid structures were predicted for the reference geometry of one cooling finger. To assess numerical errors, different meshes (hexagonal, tetra, tetra-prism) and discretisation schemes were used. The temperatures in the solid structures decrease with finer mesh and higher order discretisation and converge towards finite values. Numerical simulations were validated against high heat flux experiments, performed at Efremov Institute, St. Petersburg. The predicted design parameters show reasonable agreement with measured data. The calculated maximum thimble temperature was below the tile-thimble brazing temperature, indicating good heat removal capability of reference divertor design. (author)

Construction and performance tests of Helium Engineering Demonstration Loop (HENDEL) for VHTR

International Nuclear Information System (INIS)

Hishida, M.; Tanaka, T.; Shimomura, H.; Sanokawa, K.

1984-01-01

A helium engineering demonstration loop (HENDEL) was constructed and operated in JAERI in order to develop the high-temperature key components of an experimental very high temperature gas cooled reactor, like fuel stack, in-core reactor structure, hot gas duct, intermediate heat exchanger. Performance tests as well as demonstration of integrity are carried out with large-size or actual-size models of key components. The key components to be tested in HENDEL are: fuel stack and control rod; core supporting structure, or bottom structure of rector core exposed to direct impingement of high temperature core outlet flow; reactor internal components and structure; high temperature components in heat removal system (primary and secondary cooling systems)

The primary circuit of the dragon high temperature reactor experiment

International Nuclear Information System (INIS)

Simon, R.

2005-01-01

The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01

A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

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

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Reyes-Ramirez, Ricardo; Francois, Juan-Luis; Reinking-Cejudo, Arturo G.

2011-01-01

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

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.

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

Energy Technology Data Exchange (ETDEWEB)

Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

1992-02-01

This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

International Nuclear Information System (INIS)

Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

1992-02-01

This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

International Nuclear Information System (INIS)

Klein, Andrew; Lenhof, Renae; Deason, Wesley; Harter, Jackson

2015-01-01

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

Energy Technology Data Exchange (ETDEWEB)

Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

2015-01-16

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2010-07-01

SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

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.

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Gas-Cooled Reactors: the importance of their development

International Nuclear Information System (INIS)

Kasten, P.R.

1978-01-01

Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

An efficient continuous flow helium cooling unit for Moessbauer experiments

International Nuclear Information System (INIS)

Herbert, I.R.; Campbell, S.J.

1976-01-01

A Moessbauer continuous flow cooling unit for use with liquid helium over the temperature range 4.2 to 300K is described. The cooling unit can be used for either absorber or source studies in the horizontal plane and it is positioned directly on top of a helium storage vessel. The helium transfer line forms an integral part of the cooling unit and feeds directly into the storage vessel so that helium losses are kept to the minimum. The helium consumption is 0.12 l h -1 at 4.2 K decreasing to 0.055 l h -1 at 40 K. The unit is top loading and the exchange gas cooled samples can be changed easily and quickly. (author)

Computer simulation of multiple stability regions in an internally cooled superconducting conductor and of helium replenishment in a bath-cooled conductor

International Nuclear Information System (INIS)

Turner, L.R.; Shindler, J.

1984-09-01

For upcoming fusion experiments and future fusion reactors, superconducting magnetic have been chosen or considered which employ cooling by pool-boiling HeI, by HeII, and by internally flowing HeI. The choice of conductor and cooling method should be determined in part by the response of the magnet to sudden localized heat pulses of various magnitudes. The paper describes the successful computer simulation of multiple stability in internally cooled conductors, as observed experimentally, using the computer code SSICC. It also describes the modeling of helium replenishment in the cooling channels of a bath-cooled conductor, using the computer code TASS

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

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

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

Application of Hastelloy X in gas-cooled reactor systems

International Nuclear Information System (INIS)

Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

1976-10-01

Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

Science.gov (United States)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-01-01

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

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

Uranium utilization of light water cooled reactors and fast breeders

International Nuclear Information System (INIS)

Stojadinovic, Timm

1991-08-01

The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

Liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Leigh, K.M.

1980-01-01

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

High temperature gas-cooled reactors - once-through fuel cycle

International Nuclear Information System (INIS)

1979-03-01

The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U 3 O 8 requirement of 4280 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U 3 O 8 requirement by about 9%. Additional resource utilization improvements of 10% could be realized by improved fuel management techniques. In addition to improvements achieved in reactor technology, uranium utilization can also be improved by reducing the U-235 content in the depleted uranium (tails) produced by the isotope separation facility. If the Advanced Isotope Separation Technology program, currently under development by the United States, results in a lowering of the tails assay from 0.20 w/o to 0.05 w/o the uranium feed requirement for MEU/Th cycles would be further reduced by 22%. A total improvement of 41% over the already relatively low 4280 ST/GWe net lifetime U 3 O 8 requirement would result in a 2525 ST/GWe 30-year yet U 3 O 8 requirement if all of the potential improvements were realized

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.

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

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

The Modular Helium Reactor for Hydrogen Production

International Nuclear Information System (INIS)

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

2006-01-01

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

Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

International Nuclear Information System (INIS)

Boer, Brian

2007-01-01

The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

Cooling performance of helium-gas/water coolers in HENDEL

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

1994-01-01

The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

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

International Nuclear Information System (INIS)

Yoon, K. H.; Lee, C. B.

2014-01-01

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

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.

Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

International Nuclear Information System (INIS)

Smith, J.; Hart, R.; Lazic, L.

2009-01-01

allow steam at the wellhead to be at 8.5 MPa, saturated, despite significant steam pipe lengths. The steam generator system consists of steam generators, pre-heaters and super-heaters, all designed for operation with high temperature helium as a heat transfer medium. This design utilizes worldwide nuclear steam generator as well as fossil-fuel steam generator experience for optimized, reliable performance. The paper describes the safety aspects of the steam generator system, overall layout of the gas-cooled reactor plant and system controls. With this system, the gas-cooled reactor becomes a viable alternative for energy supply in the Oil Sands. (author)

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

Energy Technology Data Exchange (ETDEWEB)

Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

Reactor core of light water-cooled reactor

International Nuclear Information System (INIS)

Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

1996-01-01

In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

HTGR [High Temperature Gas-Cooled Reactor] ingress analysis using MINET

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

1989-04-01

Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs

On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

International Nuclear Information System (INIS)

Hawari, Ayman I.; Bourham, Mohamed A.

2010-01-01

Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Manufacturing aspects in the design of the breeder unit for Helium Cooled Pebble Bed blankets

International Nuclear Information System (INIS)

Rey, J.; Ihli, T.; Filsinger, D.; Polixa, C.

2007-01-01

The breeding blanket programme has been in the focus of European fusion research for more than a decade. Recently, it has been driven by the EU Power Plant Conceptual Study (PPCS), investigating the potential of fusion energy in a future economic environment. On the way to the first commercial nuclear fusion reactor (DEMO) new studies for reactor in-vessel components have been initiated. One central focus is the design and manufacturing of the blankets that have to ensure the breeding process to maintain the fuel cycle and are also responsible for the extraction of the main part of the reactor heat for power generation. Two kinds are established: One is the Helium Cooled Pebble Bed (HCPB) and the other the Helium Cooled Liquid Lead (HCLL) blanket. Both designs employ three different cooling plate assemblies. The outer, cooled U-shaped shell, namely the First Wall (FW), with two caps builds the blanket box. The structural strength of the blanket box is realized by integrating Stiffening Grids (SG) that separate the equally spaced Breeder Unit (BU) and allow the box, in case of faulted conditions, to withstand an internal pressure of 8 MPa. The cooled SG constitute the side walls of the BU and are also cooled. The BU consists of a dedicated Cooling Plate (CP) assembly. In present studies about the fabrication of Cooling Plates two kinds of diffusion welding processes are focused on. One is based on a Hot Isostatic Gas Process (HIP). The second is a uni-axial Diffusion Welding Process (DWP). In both cases the bond between the two halves of the cooling plate structure is reached by controlled pressure and heat cycles. Approaching larger, realistic scaled components the uncertainty of ensuring uniform process parameters across the bonding zone increases the risk of defect sources and, therefore, makes it difficult to guarantee the required bonding penetration. This study presents an alternative manufacturing strategy. The premises for this strategy are the reduction of

A robust helium-cooled shield/blanket design for ITER

International Nuclear Information System (INIS)

Wong, C.P.C.; Bourque, R.F.; Baxi, C.B.

1993-11-01

General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding, its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology

Emergency cooling system for nuclear reactors

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

[Fluid dynamics of supercritical helium within internally cooled cabled superconductors

International Nuclear Information System (INIS)

Van Sciver, S.W.

1995-01-01

The Applied Superconductivity Center of the University of Wisconsin-Madison proposes to conduct research on low temperature helium fluid dynamics as it applies to the cooling of internally cooled cabled superconductors (ICCS). Such conductors are used in fusion reactor designs including most of the coils in ITER. The proposed work is primarily experimental involving measurements of transient and steady state pressure drop in a variety of conductor configurations. Both model and prototype conductors for actual magnet designs will be investigated. The primary goal will be to measure and model the friction factor for these complex geometries. In addition, an effort will be made to study transient processes such as heat transfer and fluid expulsion associated with quench conditions

A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

International Nuclear Information System (INIS)

Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

2016-01-01

The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

Status and prospects for gas cooled reactor fuels. Proceedings of two IAEA meetings held in June 2004 and June 2005

International Nuclear Information System (INIS)

2009-04-01

Recently, efforts to develop high temperature gas cooled reactors with an aim to building futuristic nuclear energy systems with advanced nuclear fuel cycles in the context of the Generation IV International Forum have increased significantly. In addition, several development projects are ongoing, focusing on the burning of weapons grade plutonium, including civil plutonium and other transuranic elements using the 'deep-burn concept', or 'inert matrix fuels', especially in the form of coated particles in gas cooled reactor systems. There is also considerable global interest in developing 'nuclear hydrogen' energy systems using high temperature gas cooled reactors. Apart from these developments, the value of preserving the large technology base developed in Germany, the United Kingdom and the United States of America, as well as information developed in other countries, has also been a subject of interest to the IAEA. At the second annual meeting of the 'technical working group on nuclear fuel cycles options and spent fuel management' (TWG-NFCO), held in Vienna from 28-30 May 2003, it was recommended to hold a technical meeting on Current Status and Future Prospects of Gas Cooled Reactor Fuels. The meeting should cover the technological progress that has been made in the last three years and plan future fabrication and qualification facilities for GCR/HTR fuel. TWG-NFCO considered it timely that this progress should be presented and discussed in the interested community. Recognizing the numerous activities being pursued in many Member States, the IAEA convened the technical meeting on this topic in June 2004 in Vienna. Consequently, an update meeting was held in June 2005, which was hosted by the Kharkov Institute of Physics and Technology of Ukraine to review and integrate the latest developments. This publication combines the results of the technical meeting of June 2004 and the meeting of June 2005. The proceedings presented here contain 25 in depth papers on the

Effects of cooling time on a closed LWR fuel cycle

International Nuclear Information System (INIS)

Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

2012-01-01

In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

The modular high-temperature gas-cooled reactor (MHTGR) in the US

International Nuclear Information System (INIS)

Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

1987-01-01

GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

International Nuclear Information System (INIS)

Parks, G.T.

1987-01-01

A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

Improvements in or relating to cooling systems for nuclear fuel assemblies

International Nuclear Information System (INIS)

Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

1980-01-01

A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

International Nuclear Information System (INIS)

You, WuSeung; Hong, Ser Gi

2014-01-01

The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Design study on the helium engineering demonstration loop (HENDEL)

International Nuclear Information System (INIS)

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

1977-11-01

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

Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

International Nuclear Information System (INIS)

Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

1992-01-01

In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

Status of and prospects for gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

IAEA high temperature gas-cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2000-01-01

The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

IAEA high temperature gas cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Directory of Open Access Journals (Sweden)

Phani Kumar Domalapally

2015-04-01

Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

International Nuclear Information System (INIS)

2012-12-01

The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

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

International Nuclear Information System (INIS)

2002-01-01

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

Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

International Nuclear Information System (INIS)

Su'ud, Zaki

2008-01-01

In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

Assessing the feasibility of a high-temperature, helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

International Nuclear Information System (INIS)

Barnard, H.S.; Hartwig, Z.S.; Olynyk, G.M.; Payne, J.E.

2012-01-01

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B 0 = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m −2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ∼1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

2009-01-01

Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

International Nuclear Information System (INIS)

Weaver, K. D.

2005-01-01

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 participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above

Research reactor de-fueling and fuel shipment

International Nuclear Information System (INIS)

Ice, R.D.; Jawdeh, E.; Strydom, J.

1998-01-01

Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

International Nuclear Information System (INIS)

Inhoven, H.

1984-12-01

KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.) [de

Natural convection in closed vertical cylinders with particular reference to gas cooled reactor standpipes

International Nuclear Information System (INIS)

Spence, I.D.

1975-09-01

The access to the core for fuel assemblies and control rods of the Advanced Gas Cooled Reactor is through the top cap by means of standpipes. The standpipe is essentially a cylindrical, vertical tube with cooled side wall, closed upper end and an orifice at the lower end which is exposed to the hot core fluid. This creates confined natural convection flow in the empty standpipe and this is the subject of this thesis. The investigation is carried out using analytical and experimental methods. For the analytical work, solution of laminar and turbulent flow is attempted using finite-difference computer techniques. The laminar flow performance is evaluated using two different finite-difference procedures, and the results are compared to each other and to existing analytical and experimental results for the open thermosyphon with cool inflow and hot sidewall, i.e. the complementary problem to the present one. For turbulent flow a two equation turbulence model is employed which provides transport equations for the kinetic energy of turbulence and its dissipation rate. The experimental rig is a full scale replica of the Advanced Gas Cooled Reactor control rod mechanism standpipe. Carbon dioxide and helium are used as the working fluids for the series of tests. (author)

The calculating methods of the release of airborne radionuclides to environment during the normal operation of a module high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liu Yuanzhong

1993-01-01

The calculations of the release of radionuclides to environment are the basis of environmental impact assessment during the normal operation of a module high temperature gas-cooled reactor of the Institute of Nuclear Energy Technology, Tsinghua University, China. According to the features of the reactor it is pointed out that only five sources of the airborne radioactive materials released to environment are important. They are: (1) the activation of the air in the reactor cavity; (2) the escape from the primary coolant systems; (3) the release of radioactively contaminated helium from storage tanks; (4) the release of radioactively contaminated helium from the gas evacuation system of fuel load and unload system; (5) the leakage of the vapour from water-steam loop. In accordance with five release sources the calculating methods of radionuclides released to environment are worked out respectively and the respective calculating formulas are derived for the normal operation of the reactor

Emergency cooling device for reactors

International Nuclear Information System (INIS)

Inoue, Hisamichi; Naito, Masanori; Sato, Chikara; Chino, Koichi.

1975-01-01

Object: To pour high pressure cooling water into a core, when coolant is lost in a boiling water reactor, thereby restraining the rise of fuel cladding. Structure: A control rod guiding pipe, which is moved up and down by a control rod, is mounted on the bottom of a pressure vessel, the control rod guiding pipe being communicated with a high pressure cooling water tank positioned externally of the pressure vessel, and a differential in pressure between the pressure vessel and the aforesaid tank is detected when trouble of coolant loss occurs, and the high pressure cooling water within the tank is poured into the core through the control rod guiding pipe to restrain the rise of fuel cladding. (Kamimura, M.)

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

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

International Nuclear Information System (INIS)

Baldwin, D.H.

1984-08-01

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

Passive cooling of a fixed bed nuclear reactor

International Nuclear Information System (INIS)

Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

2005-01-01

Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

International Nuclear Information System (INIS)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

2015-01-01

Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO 2 emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO 2 emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus TM Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition

Optimization of the fuel assembly for the Canadian SuperCritical Water-cooled Reactor (SCWR)

Energy Technology Data Exchange (ETDEWEB)

French, C., E-mail: Corey.French@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada); Bonin, H.; Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

2013-07-01

An approach to develop a parametric optimization tool to support the Canadian Supercritical Water-cooled Reactor (SCWR) fuel design is presented in this work. The 2D benchmark lattices for 78-pin and 64-pin fuel assemblies are used as the initial models from which fuel performance and subsequent optimization stem from. A tandem optimization procedure is integrated which employs the steepest descent method. The physics codes WIMS-AECL, MCNP6 and SERPENT are used to calculate and verify select performance factors. The results are used as inputs to an optimization algorithm that yield optimal fresh fuel isotopic composition and lattice geometry. Preliminary results on verifications of infinite lattice reactivity are demonstrated in this paper. (author)

Method of operating a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Lysell, G.

1975-01-01

When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

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.

Specific power of liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Dobranich, D.

1987-10-01

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

Liquid metal cooled fast breeder nuclear reactor

International Nuclear Information System (INIS)

Scott, D.

1979-01-01

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

Gas cooled thermal reactors with high temperatures (VHTR)

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.

2014-01-01

VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)

Strength analysis of CARR-CNS with crescent-shape moderator cell and helium sub-cooling jacket covering cell

International Nuclear Information System (INIS)

Yu Qingfeng; Feng Quanke; Kawai Takeshi; Shen Feng; Yuan Luzheng; Cheng Liang

2005-01-01

The new type of the moderator cell was developed for the cold neutron source (CNS) of the China Advanced Research Reactor (CARR) which is now being constructed at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the helium sub-cooling jacket is adopted. The structure of the moderator cell is optimized by the stress FEM analysis. A crescent-shape would help to increase the volume of the moderator cell for fitting it to the four cold neutron guide tubes, even if liquid hydrogen, not liquid deuterium, was used as a cold moderator. The helium sub-cooling jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the outer shell of the moderator cell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down first into the helium sub-cooling jacket and then flows up to the condenser. The theory of the self-regulation suitable to the thermo-siphon type of the CNS is also applicable and validated

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

Science.gov (United States)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

International Nuclear Information System (INIS)

Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

1998-11-01

The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

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

International Nuclear Information System (INIS)

Dumm, K.; Donaldson, J.

1988-01-01

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

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

International Nuclear Information System (INIS)

Hewing, G.

1977-01-01

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

Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

International Nuclear Information System (INIS)

Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

2014-01-01

A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

Characterization of effluents from a high-temperature gas-cooled reactor fuel refabrication plant

International Nuclear Information System (INIS)

Judd, M.S.; Bradley, R.A.; Olsen, A.R.

1975-12-01

The types and quantities of chemical and radioactive effluents that would be released from a reference fuel refabrication facility for the High-Temperature Gas-Cooled Reactor (HTGR) have been determined. This information will be used to predict the impact of such a facility on the environment, to identify areas where additional development work needs to be done to further identify and quantify effluent streams, and to limit effluent release to the environment

Study on plant concept for gas cooled fast reactor

International Nuclear Information System (INIS)

Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

2003-05-01

In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-15

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-01

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

Reactor fuel charging equipment

International Nuclear Information System (INIS)

Wade, Elman.

1977-01-01

In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

International Nuclear Information System (INIS)

Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

1976-11-01

During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C 2 Cl 4 ) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed

RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

International Nuclear Information System (INIS)

Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

2004-01-01

RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

2003-02-01

This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Schluderberg, D.C.

1981-01-01

A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

Vented fuel experiment for gas-cooled fast reactor application

International Nuclear Information System (INIS)

Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

1976-01-01

A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

Hydrogen Process Coupling to Modular Helium Reactors

International Nuclear Information System (INIS)

Shenoy, Arkal; Richards, Matt; Buckingham, Robert

2009-01-01

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

The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

International Nuclear Information System (INIS)

Mears, L.D.; Dean, R.A.

1987-01-01

The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

Performance of highly rated UO₂ fuel in the WR-1 organic-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Schankula, M. H.; Hastings, I. J.

1977-07-15

Information on oxide fuel behaviour in organic coolant was required as part of the organic-cooled power reactor (OCR) study. Of major interest were data on the release of fission gases from fuel operating at high fuel surface temperatures and low external restraint; features which are peculiar to the OCR. To provide these and other data, UO₂ fuel with cold-worked Zr-2.5wt%Nb sheathing was irradiated in the WR-1 organic-cooled reactor to burnups of 135-154 MWh/kgU at a time-averaged linear power of 60-63 kW/m. Elements with 0.38 and 0.69 mm thick sheathing showed maximum diametral increases averaging 3.7 and 1.7% respectively at pellet mid-planes. Reduced fuel/sheath heat transfer resulting from a difference between internal gas pressure and coolant pressure produced high operating temperatures, and there was evidence of central melting in some elements. Fission gas releases were 30-60%. In the heat affected zone adjacent to brazed appendages, the diametral increases were lower, averaging 0.9 and 0.5% for 0.38 and 0.69 mm thick sheathing respectively. Heat treatment during the brazing process produced a local improvement in sheath creep strength. Highly rated oxide fuel irradiated in organic coolant will require sheathing with improved high temperature creep properties; heat-treated Zr-2.5 wt% Nb may provide this improvement.

Design, fabrication, and testing of a helium-cooled module for the ITER divertor

International Nuclear Information System (INIS)

Baxi, C.B.; Smith, J.P.; Youchison, D.

1994-08-01

The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

Nuclear fuel rod helium leak inspection apparatus and method

International Nuclear Information System (INIS)

Ahmed, H.J.

1991-01-01

This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod