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Sample records for htgr fuel particle

  1. HTGR fuel particle crusher design evaluation

    International Nuclear Information System (INIS)

    Johanson, N.W.

    1978-10-01

    This report describes an evaluation of the design of the existing engineering-scale fuel particle crushing system for the HTGR reprocessing cold pilot plant at General Atomic Company (GA). The purpose of this evaluation is to assess the suitability of the existing design as a prototype of the HTGR Recycle Reference Facility (HRRF) particle crushing system and to recommend alternatives where the existing design is thought to be unsuitable as a prototype. This evaluation has led to recommendations for an upgraded design incorporating improvements in bearing and seal arrangement, housing construction, and control of roll gap thermal expansion. 23 figures, 6 tables

  2. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power

  3. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power.

  4. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  5. Automatic particle-size analysis of HTGR recycle fuel

    International Nuclear Information System (INIS)

    Mack, J.E.; Pechin, W.H.

    1977-09-01

    An automatic particle-size analyzer was designed, fabricated, tested, and put into operation measuring and counting HTGR recycle fuel particles. The particle-size analyzer can be used for particles in all stages of fabrication, from the loaded, uncarbonized weak acid resin up to fully-coated Biso or Triso particles. The device handles microspheres in the range of 300 to 1000 μm at rates up to 2000 per minute, measuring the diameter of each particle to determine the size distribution of the sample, and simultaneously determining the total number of particles. 10 figures

  6. Crushing strength of HTGR fuel particles

    International Nuclear Information System (INIS)

    Lackey, W.J.; Stinton, D.P.; Davis, L.E.; Beatty, R.L.

    1976-01-01

    The whole-particle crushing strengths of High-Temperature Gas-Cooled Reactor fertile and fissile coated particles were measured and correlated with fabrication procedures. The crushing strength of Biso-coated fertile particles was increased by the following factors: (1) increasing the outer coating thickness by 10 μm increased strengths by 0.3 lb (1.3 N) for annealed particles and by 0.5 lb (2.2 N) for unannealed particles. (2) An 1800 0 C postcoating anneal increased strengths by 1 lb (4.4 N) for particles with thick outer coatings and by 2 lb (8.9 N) for particles having thin coatings. (3) Increasing the inner coating density by 0.1 g/cm 3 increased strength by 0.6 lb (2.7 N). The crushing strength of Triso-coated fissile particles was proportional to the thickness of the SiC coatings, and strength decreased on annealing by about 0.2 lb (0.9 N) when a porous plate was used to distribute the coating gas and by about 1.5 lb (6.7 N) when a conical gas distributor was used. The strengths of fertile and fissile coated particles as well as uncoated kernels appear adequate to allow fuel fabrication without excessive particle damage

  7. Evaluation of a blender for HTGR fuel particles

    International Nuclear Information System (INIS)

    Johnson, D.R.

    1977-03-01

    An experimental blender for mixing HTGR fuel particles prior to molding the particles into fuel rods was evaluated. The blender consists of a conical chamber with an air inlet in the bottom. A pneumatically operated valve provides for discharge of the particles out of the bottom of the cone. The particles are mixed by periodically levitating with pulses of air. The blender has provision for regulating the air flow rate and the number and duration of the air flow pulses. The performance of the blender was governed by the particle blend being mixed, the air flow rate, and the pulse time. Adequately blended fuel rods can be made, if the air flow rate and pulse time are carefully controlled for each fuel rod composition

  8. Chemical thermodynamics of iodine species in the HTGR fuel particle

    International Nuclear Information System (INIS)

    Lindemer, T.B.

    1982-09-01

    The iodine-containing species in an intact fuel particle in the high-temperature gas-cooled reactor (HTGR) have been calculated. Assumptions include: (1) attainment of chemical thermodynamic equilibrium among all species in the open porosity of the particle, primarily in the buffer layer; and (2) fission-product concentrations in proportion to their yields. The primary gaseous species is calculated to be cesium iodide; in carbide-containing fuels, gaseous barium iodide may exhibit equivalent pressures. The condensed iodine-containing phase is usually cesium iodide, but in carbide-containing fuels, barium iodide may be stable instead. Absorption of elemental iodine on the carbon in the particle appears to be less than or equal to 10 -4 μg I/g C. The fission-product-spectra excess of cesium over iodine would generally be adsorbed on the carbon, but may form Cs 2 MoO 4 under some circumstances

  9. Device for sampling HTGR recycle fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Lackey, W.J.

    1977-03-01

    Devices for sampling High-Temperature Gas-Cooled Reactor fuel microspheres were evaluated. Analysis of samples obtained with each of two specially designed passive samplers were compared with data generated by more common techniques. A ten-stage two-way sampler was found to produce a representative sample with a constant batch-to-sample ratio

  10. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  11. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  12. On transient irradiation behavior of HTGR fuel particles

    International Nuclear Information System (INIS)

    Mortenson, S.C.; Okrent, D.

    1977-01-01

    An examination of HTGR TRISO coated fuel particles was made in which the particles' stress-strain histories were determined during both steady-state and transient operating conditions. The basis for the examination was a modified version of a computer code written by Kaae which assumed spherical symmetry, isotropic thermal expansion, isotropic elastic constants, time-temperature-irradiation invariant materials properties, and steady state operation during particle exposure. Additionally, the Kaae code modelled potential separation of layers at the SiC-inner PyC interface and considered that several entrapped fission products could exist in either the gaseous or solid state, dependent upon particle operating conditions. Using the modified code which modelled transient behavior in a quasi-static fashion, a series of both steady-state and transient operating condition computer simulations was made. For the former set of runs, a candidate set of particle dimensions and a nominal set of materials' properties was assumed. Layer thicknesses were assumed to be normally distributed about the nominal thickenesses and a probability distribution of SiC tensile stresses was generated; sensitivity of the stress distribution to assumed standard deviation of the layer thicknesses was acute. Further, this series of steady-state runs demonstrated that for certain combinations of the assumed PyC-SiC bond interface strength and irradiation-induced creep constant, anomalous predicted stresses may be obtained in the PyC layers. The steady-state runs also suggest that transient behavior would most likely not be significant at fast neutron exposures below about 10 21 NVT due to both low fission gas pressure and likely beneficial interface separation

  13. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  14. HTGR fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-01-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  15. Factors affecting defective fraction of biso-coated HTGR fuel particles during in-block carbonization

    International Nuclear Information System (INIS)

    Caputo, A.J.; Johnson, D.R.; Bayne, C.K.

    1977-01-01

    The performance of Biso-coated thoria fuel particles during the in-block processing step of HTGR fuel element refabrication was evaluated. The effect of various process variables (heating rate, particle crushing strength, horizontal and/or vertical position in the fuel element blocks, and fuel hole permeability) on pitch coke yield, defective fraction of fuel particles, matrix structure, and matrix porosity was evaluated. Of the variables tested, only heating rate had a significant effect on pitch coke yield while both heating rate and particle crushing strength had a significant effect on defective fraction of fuel particles

  16. Development of a pneumatic transfer system for HTGR recycle fuel particles

    International Nuclear Information System (INIS)

    Mack, J.E.; Johnson, D.R.

    1978-02-01

    In support of the High-Temperature Gas-Cooled Reactor (HTGR) Fuel Refabrication Development Program, an experimental pneumatic transfer system was constructed to determine the feasibility of pneumatically conveying pyrocarbon-coated fuel particles of Triso and Biso designs. Tests were conducted with these particles in each of their nonpyrophoric forms to determine pressure drops, particle velocities, and gas flow requirements during pneumatic transfer as well as to evaluate particle wear and breakage. Results indicated that the material can be pneumatically conveyed at low pressures without excessive damage to the particles or their coatings

  17. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.; Lindemer, T.B.; Beatty, R.L.; Tiegs, T.N.

    1976-12-01

    Recycle fissile fuel particles for high-temperature gas-cooled reactors (HTGRs) have been under development since the mid-1960s. Irradiation performance on early UO 2 and Th 0 . 8 U 0 . 2 O 2 kernels is described in this report, and the performance limitations associated with the dense oxide kernels are presented. The development of the new reference fuel kernel, the weak-acid-resin-derived (WAR) UO 2 --UC 2 , is discussed in detail, including an extensive section on the irradiation performance of this fuel in HFIR removable beryllium capsules HRB-7 through -10. The conclusion is reached that the irradiation performance of the WAR fissile fuel kernel is better than that of any coated particle fuel yet tested. Further, the present fissile kernel is adequate for steam cycle HTGRs as well as for many advanced applications such as gas turbine and process heat HTGRs

  18. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  19. Fission product behavior in HTGR fuel particles made from weak-acid resins

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Henson, T.J.

    1979-04-01

    Fission product retention and behavior are of utmost importance in HTGR fuel particles. The present study concentrates on particles made from weak-acid resins, which can vary in composition from 100% UO 2 plus excess carbon to 100% UC 2 plus excess carbon. Five compositions were tested: UC 4 58 O 2 04 , UC 3 68 O 0 01 , UC 4 39 O 1 72 , UC 4 63 O 0 97 , and UC 4 14 O 1 53 . Metallographically sectioned particles were examined with a shielded electron microprobe. The distributions of the fission products were determined by monitoring characteristic x-ray lines while scanning the electron beam over the particle surface

  20. Automatic particle-size analysis of HTGR nuclear fuel microspheres

    International Nuclear Information System (INIS)

    Mack, J.E.

    1977-01-01

    An automatic particle-size analyzer (PSA) has been developed at ORNL for measuring and counting samples of nuclear fuel microspheres in the diameter range of 300 to 1000 μm at rates in excess of 2000 particles per minute, requiring no sample preparation. A light blockage technique is used in conjunction with a particle singularizer. Each particle in the sample is sized, and the information is accumulated by a multi-channel pulse height analyzer. The data are then transferred automatically to a computer for calculation of mean diameter, standard deviation, kurtosis, and skewness of the distribution. Entering the sample weight and pre-coating data permits calculation of particle density and the mean coating thickness and density. Following this nondestructive analysis, the sample is collected and returned to the process line or used for further analysis. The device has potential as an on-line quality control device in processes dealing with spherical or near-spherical particles where rapid analysis is required for process control

  1. Fluidized combustion of beds of large, dense particles in reprocessing HTGR fuel

    International Nuclear Information System (INIS)

    Young, D.T.

    1977-03-01

    Fluidized bed combustion of graphite fuel elements and carbon external to fuel particles is required in reprocessing high-temperature gas-cooled reactor (HTGR) cores for recovery of uranium. This burning process requires combustion of beds containing both large particles and very dense particles as well as combustion of fine graphite particles which elutriate from the bed. Equipment must be designed for optimum simplicity and reliability as ultimate operation will occur in a limited access ''hot cell'' environment. Results reported in this paper indicate that successful long-term operation of fuel element burning with complete combustion of all graphite fines leading to a fuel particle product containing <1% external carbon can be performed on equipment developed in this program

  2. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

    In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL) [de

  3. Improved gas distributor for coating HTGR fuel particles

    International Nuclear Information System (INIS)

    Lackey, W.J.; Stinton, D.P.; Sease, J.D.

    1977-01-01

    The important criteria to be considered in design of the gas distributor are: (1) The distributor should ideally spread or disperse the gas over the full area of the coating chamber to maximize the particle gas contact area and thereby increase both particle circulation and the percentage of the input gas that ends up as coating. (2) The gas should not heat up during its passage through the distributor. Otherwise the gas would partially decompose prematurely, causing excessive coating deposition within or on the distributor. (3) The distributor should be designed to minimize accidental drainage of particles from the furnace and blowover of particles into the effluent system. (4) The distributor should be capable of depositing both carbon and SiC coatings of high quality as regards to density, preferred orientation, permeability, defective fraction, and other product attributes. (5) The distributor should be amenable to use with large particle charges and short turnaround times and be simple, inexpensive, and reliable. We have devised a simple distributor that incorporates the five criteria listed above. The new design is termed a blind-hole frit. All the gas passes through the thinned blind-hole regions, and thus the gas velocity is considerably higher than for a flat frit of uniform thickness. Because of its high velocity, the gas does not have time to reach a high enough temperature to cause deposition within the frit. Also most of the resistance to gas flow is provided by the porous distributor and not by the particle bed; therefore, localized variations of the quantity of particles above any particular gas inlet do not significantly alter the flow rate through that inlet

  4. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

    1985-01-01

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  5. Irradiated-Microsphere Gamma Analyzer (IMGA): an integrated system for HTGR coated particle fuel performance assessment

    International Nuclear Information System (INIS)

    Kania, M.J.; Valentine, K.H.

    1980-02-01

    The Irradiated-Microsphere Gamma Analyzer (IMGA) System, designed and built at ORNL, provides the capability of making statistically accurate failure fraction measurements on irradiated HTGR coated particle fuel. The IMGA records the gamma-ray energy spectra from fuel particles and performs quantitative analyses on these spectra; then, using chemical and physical properties of the gamma emitters it makes a failed-nonfailed decision concerning the ability of the coatings to retain fission products. Actual retention characteristics for the coatings are determined by measuring activity ratios for certain gamma emitters such as 137 Cs/ 95 Zr and 144 Ce/ 95 Zr for metallic fission product retention and 134 Cs/ 137 Cs for an indirect measure of gaseous fission product retention. Data from IMGA (which can be put in the form of n failures observed in N examinations) can be accurately described by the binomial probability distribution model. Using this model, a mathematical relationship between IMGA data (n,N), failure fraction, and confidence level was developed. To determine failure fractions of less than or equal to 1% at confidence levels near 95%, this model dictates that from several hundred to several thousand particles must be examined. The automated particle handler of the IMGA system provides this capability. As a demonstration of failure fraction determination, fuel rod C-3-1 from the OF-2 irradiation capsule was analyzed and failure fraction statistics were applied. Results showed that at the 1% failure fraction level, with a 95% confidence level, the fissile particle batch could not meet requirements; however, the fertile particle exceeded these requirements for the given irradiation temperature and burnup

  6. Influence of process variables on permeability and anisotropy of Biso-coated HTGR fuel particles

    International Nuclear Information System (INIS)

    Stinton, D.P.; Lackey, W.J.; Thiele, B.A.

    1977-11-01

    The effect of several important process variables on the fraction of defective particles and anisotropy of the low-temperature isotropic (LTI) coating layer was determined for Biso-coated HTGR fuel particles. Process variables considered are deposition temperature, hydrocarbon type, diluent type, and percent diluent. The effect of several other variables such as coating rate and density that depend on the process variables were also considered in this analysis. The fraction of defective particles was controlled by the dependent variables coating rate and LTI density. Coating rate was also the variable controlling the anisotropy of the LTI layer. Diluent type and diluent concentration had only a small influence on the deposition rate of the LTI layer. High-quality particles in terms of anisotropy and permeability can be produced by use of a porous plate gas distributor if the coating rate is between 3 and 5 μm/min and the coating density is between about 1.75 and 1.95 g/cm 3

  7. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  8. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

    An overview of HTGR fuel recycle is presented, with emphasis placed on reprocessing and fuel kernel refabrication. Overall recycle operations include (1) shipment and storage, (2) reprocessing, (3) refabrication, (4) waste handling, and (5) accountability and safeguards

  9. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  10. In-line monitoring of effluents from HTGR fuel particle preparation processes using a time-of-flight mass spectrometer

    International Nuclear Information System (INIS)

    Lee, D.A.; Costanzo, D.A.; Stinton, D.P.; Carpenter, J.A.; Rainey, W.T. Jr.; Canada, D.C.; Carter, J.A.

    1976-08-01

    The carbonization, conversion, and coating processes in the manufacture of HTGR fuel particles have been studied with the use of a time-of-flight mass spectrometer. Non-condensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes which have been monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO 2 to UC 2 , buffer and low temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures

  11. Determination of end-of-life-failure fractions of HTGR-fuel particles by postirradiation annealing and beta autoradiography

    International Nuclear Information System (INIS)

    Thiele, B.A.; Herren, M.

    1978-11-01

    Fission-product contamination of the helium coolant of High-Temperature Gas-Cooled Reactors (HTGR) is strongly influenced by the end-of-life (EOL) failed-particle fraction. Knowledge of the EOL-failure fraction is the basis for model calculations to predict the total fission product release from the reactor core. After disintegration of irradiation fuel rods, fuel particles are placed in individual holes of a graphite tray. During a 5-h heat treatment at 1000 0 C in a helium atmosphere failed particles leak fission products, especially the volatile cesium, into the graphite. After unloading a β-autoradiograph of the tray is made. Holes that housed defective particles are identified from black spots on the β-sensitive film. The EOL-failure fraction is the ratio of defective particles to the total number of particles tested. The technique is called PIAA, PostIrradiation Annealing and Autoradiography. The PIAA technique was applied to particles of a Trisocoated highly-enriched UO 2 fissile batch irradiated to a burnup of 35% FIMA at an irradiation temperature of 1250 0 C. Visual examination showed all particles to be intact. From 11 to 47% of the particles had failed, as determined by PIAA. Further, postirradiation examination showed that localized corrosion of the silicon carbide coating by fission-product rare-earth chlorides had occurred

  12. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  13. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  14. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  15. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Baxter, B.; Benedict, G.E.; Zimmerman, R.D.

    1976-01-01

    Development studies to date indicate that the HTGR fuel blocks can be effectively crushed with two stages of eccentric jaw crushing, followed by a double-roll crusher, a screener and an eccentrically mounted single-roll crusher for oversize particles. Burner development results indicate successful long-term operation of both the primary and secondary fluidized-bed combustion systems can be performed with the equipment developed in this program. Aqueous separation development activities have centered on adapting known Acid-Thorex processing technology to the HTGR reprocessing task. Significant progress has been made on dissolution of burner ash, solvent extraction feed preparation, slurry transfer, solids drying and solvent extraction equipment and flowsheet requirements

  16. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    International Nuclear Information System (INIS)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-01

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

  17. Behaviour of HTGR coated fuel particles at high-temperature tests

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Lyutikov, R.A.; Kurbakov, S.D.; Repnikov, V.M.; Khromonozhkin, V.V.; Soloviyov, G.I.

    1990-01-01

    At the temperature range 1200-2600 deg. C prereactor tests of TRISO fuel particles on the base of UO 2 , UC x O y and UO 2 +2Al 2 O 3 . SiO 2 kernels, and also fuel particle models with ZrC kernels were performed. Isothermal annealings carried out at temperatures of 1400-2600 deg. C, thermogradient ones at 1200-2200 deg. C (Δ T = 200-1200 deg. C/cm). It is shown that at heating to 2200 deg. C integrity of fuel particles is limited by different thermal expansion of PyC and SiC coatings, and also by thermal dissociation of SiC. At higher temperatures the failure is caused by development of high pressures within weakened fuel particles. It is found that uranium migration from alloyed fuel (UC x O y , UO 2 +2Al 2 O 3 .SiO 2 ) in the process of annealing is higher than that from UO 2 . (author)

  18. Development of a FE Model for the Stress Analysis of HTGR TRISO-coated particle fuel

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Chang, J. H.

    2005-12-01

    Finite element modelling of the stresses in TRISO-coated fuel particle under normal operating conditions was carried out with use of the structural analysis computer code ABAQUS. The FE model took into account the irradiation induced swelling and the creep of the PyC layers, the internal fission gas pressure that builds up during irradiation and the constant external ambient pressure. All of the inputs such as particle dimensions, swelling rates and creep rates of PyC layers and other mechanical properties used in these calculations were adopted from Miller's publication published in 1993. The FE model was verified against Miller's solution. Results of this model were found to be in good agreement with Miller's results. With use of the FE model, the static behavior of the TRISO-coated fuel particle, such as load shares, stress contours, stress variations as a function of fluence and shape changes of the TRISO -coated layers were investigated

  19. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    Energy Technology Data Exchange (ETDEWEB)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-15

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl2Si2O8, BaAl2Si2O8and CsAlSi2O6 in the additional alumina-silica phase of the kernel.

  20. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  1. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  2. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  3. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  4. Study on the inspection item and inspection method of HTGR fuel

    International Nuclear Information System (INIS)

    Na, Sang Ho; Kim, Y. K.; Jeong, K. C.; Oh, S. C.; Cho, M. S.; Kim, Y. M.; Lee, Y. W.

    2006-01-01

    The type of HTGR(High Temperature Gas-cooled Reactor) fuel is different according to the reactor type. Generally the HTGR fuel has two types. One is a block type, which is manufactured in Japan or America. And the other is a pebble type, which is manufactured in China. Regardless of the fuel type, the fuel manufacturing process started from the coated particle, which is consisted of fuel kernel and the 4 coating layers. Korea has a plan to fabricate a HTGR fuel in near future. The appropriate quality inspection standards are requested to produce a sound and reliable coated particle for HTGR fuel. Therefore, the inspection items and the inspection methods of HTGR fuel between Japan and China, which countries have the manufacturing process, are investigated to establish a proper inspection standards of our product characteristics

  5. The calculation - experimental investigations of the HTGR fuel element construction

    International Nuclear Information System (INIS)

    Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

    1985-01-01

    One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

  6. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

    Directory of Open Access Journals (Sweden)

    Gongyi Yu

    2017-01-01

    Full Text Available A nonuniform distribution of tristructural isotropic (TRISO particles within a high-temperature gas-cooled reactor (HTGR pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT: first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

  7. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  8. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  9. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  10. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  11. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1987-01-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

  12. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  13. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  14. Fractional release of short-lived noble gases and iodine from HTGR fuel compact containing a fraction of coated fuel particles with through-coating defects

    International Nuclear Information System (INIS)

    Ogawa, Toru; Fukuda, Kosaku; Kobayashi, Fumiaki; Kikuchi, Teruo; Tobita, Tsutomu; Kashimura, Satoru; Kikuchi, Hironobu; Yamamoto, Katsumune.

    1986-10-01

    Fractional release (R/B) data of short-lived noble gases and iodine from sweep-gas irradiated HTGR fuel compacts were analyzed. Empirical formulas to predict R/B of 88 Kr as a function of temperature and fraction through-coating defects, and to calculate ratios of R/B's of other shortlived gases to that of 88 Kr were proposed. A method to predict R/B of iodine was also proposed. As for 131 I, a fission product of major safety concern, (R/B) I 131 ≅ (R/B) Xe 133 was predicted. Applying those methods, R/B from OGL-1 fuel element (5th and 6th) was predicted to show a good agreement with observation. (author)

  15. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Nick Soelberg

    2010-08-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  16. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Soelberg, Nick

    2010-01-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR 'full recycle' service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the 'pebble bed' approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R and D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in 'limited separation' or 'minimum fuel treatment' separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  17. Status of reprocessing technology in the HTGR fuel cycle

    International Nuclear Information System (INIS)

    Kaiser, G.; Merz, E.; Zimmer, E.

    1977-01-01

    For more than ten years extensive R and D work has been carried out in the Federal Republic of Germany in order to develop the technology necessary for closing the fuel cycle of high-temperature gas-cooled reactors. The efforts are concentrated primarily on fuel elements having either highly enriched 235 U or recycled 233 U as the fissile and thorium as the fertile material embedded in a graphite matrix. They include the development of processes and equipment for reprocessing and remote preparation of coated microspheres from the recovered uranium. The paper reviews the issues and problems associated with the requirements to deal with high burn-up fuel from HTGR's of different design and composition. It is anticipated that a grind-burn-leach head-end treatment and a modified THOREX-type chemical processing are the optimum choice for the flowsheet. An overview of the present status achieved in construction of a small reprocessing facility, called JUPITER, is presented. It includes a discussion of problems which have already been solved and which have still to be solved like the treatment of feed/breed particle systems and for minimizing environmental impacts envisaged with a HTGR fuel cycle technology. Also discussed is the present status of remote fuel kernel fabrication and coating technology. Additional activities include the design of a mock-up prototype burning head-end facility, called VENUS, with a throughput equivalent to about 6000 MW installed electrical power, as well as a preliminary study for the utilisation of the Karlsruhe LWR prototype reprocessing plant (WAK) to handle HTGR fuel after remodelling of the installations. The paper concludes with an outlook of projects for the future

  18. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

  19. Calorimetric assay of HTGR fuel samples

    International Nuclear Information System (INIS)

    Allen, E.J.; McNeany, S.R.; Jenkins, J.D.

    1979-04-01

    A calorimeter using a neutron source was designed and fabricated by Mound Laboratory, according to ORNL specifications. A calibration curve of the device for HTGR standard fuel rods was experimentally determined. The precision of a single measurement at the 95% confidence level was estimated to be +-0.8 μW. For a fuel sample containing 0.3 g 235 U and a neutron source containing 691 μg 252 Cf, this represents a relative standard deviation of 0.5%. Measurement time was approximately 5.5 h per sample. Use of the calorimeter is limited by its relatively poor precision, long measurement time, manual sample changing, sensitivity to room environment, and possibility of accumulated dust blocking water flow through the calorimeter. The calorimeter could be redesigned to resolve most of these difficulties, but not without significant development work

  20. Characteristics of radioactive waste streams generated in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Lin, K.H.

    1976-01-01

    Results are presented of a study concerned with identification and characterization of radioactive waste streams from an HTGR fuel reprocessing plant. Approximate quantities of individual waste streams as well as pertinent characteristics of selected streams have been estimated. Most of the waste streams are unique to HTGR fuel reprocessing. However, waste streams from the solvent extraction system and from the plant facilities do not differ greatly from the corresponding LWR fuel reprocessing wastes

  1. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

    International Nuclear Information System (INIS)

    Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M.; Wagner, P.

    1980-06-01

    Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer

  2. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

    Reprocessing of HTGR fuels requires evaporative concentration of uranium and thorium nitrate solutions. The results of a bench-scale test program conducted to assess the safety aspects of planned concentrator operations are reported

  3. Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

    International Nuclear Information System (INIS)

    Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

    1983-05-01

    This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

  4. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  5. Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

    International Nuclear Information System (INIS)

    Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

    1990-01-01

    In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

  6. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  7. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  8. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10 25 neutrons/m 2 to 8.5 x 10 25 neutrons/m 2 . Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating

  9. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Bayne, C.K.

    1975-01-01

    Fuel rods for the high temperature gas cooled reactor are composed of pyrolytic carbon coated fuel particles bounded by a carbonaceous matrix. Because of differential shrinkage between coated particles and the carbonaceous matrix, breakage of the pyrolytic coating has been observed with certain combinations of coated particles and matrix compositions. The pyrolytic coating is intended to be the primary containment for fission products. Therefore, an experiment is desired to determine the breakage characteristics of different strength coated particles combined with different matrix compositions during irradiation

  10. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  11. HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

    International Nuclear Information System (INIS)

    Strand, J.B.; Fields, D.E.; Kergis, C.A.

    1979-05-01

    The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

  12. HTGR fuel behavior at very high temperature

    International Nuclear Information System (INIS)

    Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

    1986-03-01

    Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

  13. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  14. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  15. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented

  16. Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y., E-mail: fukaya.yuji@jaea.go.jp; Goto, M.; Nishihara, T.

    2015-11-15

    Highlights: • We attempted and optimized erbium loading methods to improve reactivity coefficients for LRSF-HTGR. • We elucidated the mechanism of the improvements for each erbium loading method by using the Bondarenko approach. • We concluded the erbium loading method by embedding into graphite shaft is preferable. - Abstract: Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact. It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach. As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

  17. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

    International Nuclear Information System (INIS)

    Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.; Tiegs, T.N.; Montgomery, B.H.; Hamner, R.L.; Beatty, R.L.

    1977-05-01

    The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500 0 C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coated in production-scale coaters, comparison of the performance of 233 U-bearing particles with that of 235 U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of 233 U-bearing particles to be identical to that of 235 U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention

  18. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  19. Process control of an HTGR fuel reprocessing cold pilot plant

    International Nuclear Information System (INIS)

    Rode, J.S.

    1976-10-01

    Development of engineering-scale systems for a large-scale HTGR fuel reprocessing demonstration facility is currently underway in a cold pilot plant. These systems include two fluidized-bed burners, which remove the graphite (carbon) matrix from the crushed HTGR fuel by high temperature (900 0 C) oxidation. The burners are controlled by a digital process controller with an all analog input/output interface which has been in use since March, 1976. The advantages of such a control system to a pilot plant operation can be summarized as follows: (1) Control loop functions and configurations can be changed easily; (2) control constants, alarm limits, output limits, and scaling constants can be changed easily; (3) calculation of data and/or interface with a computerized information retrieval system during operation are available; (4) diagnosis of process control problems is facilitated; and (5) control panel/room space is saved

  20. Quantification of TRISO fuel heterogeneity effects in HTGR lattice physics calculations

    International Nuclear Information System (INIS)

    Perfetti, C. M.; Anghaie, S.; Dugan, E.; Marcille, T.

    2010-01-01

    A large number of LEU-MHR fuel compact models were generated with randomly distributed TRISO particle fuel and were simulated using MCNP5, and it was determined how several neutronic parameters, including k-infinite, the thermal and fast diffusion coefficients, and the four factors, varied across the randomly-generated cases. A sensitivity study was also performed to determine how the four factors depend on the definition of the thermal energy group. Values of k-infinite for the cases had a sample standard deviation of 248 pcm and were found to follow an approximately normal distribution about the mean value of k-infinite. Although all of the four factors were found to have similar sample standard deviations, the resonance escape probability was found to be the most variable parameter with a sample relative standard deviation between 0.07% and 0.08%. HTGR fuel compact homogenization methods typically examine only one reference fuel compact that contains a uniform distribution of TRISO particles, but in reality the TRISO particles are randomly distributed throughout the fuel compact. Thus, the neutronic parameters for actual fuel compacts differ randomly from those in the reference model. To license next-generation High-Temperature Gas Reactors engineers must quantify all uncertainties of the design and this random variation in neutron parameters is a previously unmeasured quantity; this study measures this uncertainty by examining the variation in k-infinite for HTGR fuel compact models with randomly distributed TRISO fuel. (authors)

  1. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  2. Calcination, Reduction and Sintering of ADU Spheres for HTGR Fuel

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Eom, Sung Ho; Kim, Yeon Ku; Kim, Woong Ki; Kim, Young Min; Lee, Young Woo; Kim, Ju Hee; Cho, Hyo Jin; Cho, Moon Seoung

    2011-01-01

    The international oil market is again in turmoil in accordance with the increasing of human needs and energy consumption. Soaring oil prices, fears of supply security, and climate change are concerned becoming more concrete make for an uncertain energy future. In this view point, nuclear energy is an important, yet controversial option for energy supply. High Temperature Gas Reactor will play a dominant role in the worldwide fleet of nuclear reactors of the next decade for electricity production and high temperature heat. HTGR have two reactor types which use the basic fuel concept based on the dispersion of TRISO coated particles in graphite in shown Fig.1. The TRISO coated particle for these purposes is prepared with pyro-carbon and silicone carbide coatings on a spherical UO 2 kernel surface as fissile material. The TRISO fuel particle consists of a microsphere (i.e., UO 2 kernel) of nuclear material: encapsulated by multiple layers of pyro-carbon and a SiC layer. This multiple coating layers system has been engineered to retain the fission products generated by fission of the nuclear material in the kernel during normal operation and all licensing basis events over the design lifetime of the fuel. UO 2 kernels are produced by using the modified sol-gel process, a wet process, generally known as the GSP method. Wet chemical processes are flexible in producing kernels of different size and chemical composition with high throughout and yield, good spherical shape, and narrow size distribution. This chemical processing route is well-known to the potential kernel fabrication processes. The principle, as set out in Fig.2, involves first of all preparing a pseudo-sol(also known as a 'broth') from an initial uranyl nitrate solution . This broth solution is obtained through addition of a number of additives, as determined by process know-how, including a soluble organic polymer, that are subsequently gels into droplets and are dispersed for ADU precipitation. The

  3. Study on the Efficient Disintegration of HTGR Fuel Elements by Electrochemical Method

    International Nuclear Information System (INIS)

    Piao Nan; Chen Ji; Xiao Cuiping; We Mingfen; Che Jing

    2014-01-01

    The spent fuel elements in High- temperature gas-cooled reactor (HTGR) have a special structure, so the head-end process of the spent fuel reprocessing is different from the process of water reactor spent fuel. The first step of head-end process of the HTGR spent fuel reprocessing process is disintegration of the graphite matrix and separation of the coated fuel particles. Electrochemical method with nitrate solution as an electrolyte for fuel element disintegration has been conducted by the Institute of Nuclear and New Energy Technology in Tsinghua University. This method allows a total disintegration of graphite matrix, while still preserving the integrity of TRISO particles. The influences of the pretreatment methods such as heating oxidation of graphite, hydrothermal and oxidants oxidation were investigated in the present work. The experimental results showed that there were no significant effects on increasing the disintegration rate when pretreatment methods were used ahead of electrochemical disintegration. This phenomenon indicated that the fuel elements which were calcined at 1073 K and pressed under 300 MPa are too compact to be broken by these pretreatment methods. And the electrochemical disintegration is an effective but slow method in breaking the graphite matrix. (author)

  4. 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)

  5. Experimental determination of thermal conductivity and gap conductance of fuel rod for HTGR

    International Nuclear Information System (INIS)

    Kikuchi, Teruo; Iwamoto, Kazumi; Ikawa, Katsuichi; Ishimoto, Kiyoshi

    1985-01-01

    The thermal conductivity of fuel compacts and the gap conductance between the fuel compact and the graphite sleeve in fuel rods for a high-temperature gas-cooled reactor (HTGR) were measured by the center heating method. These measurements were made as functions of volume percent particle loading and temperature for thermal conductivity and as functions of gap distance and gas composition for gap conductance. The thermal conductivity of fuel compacts decreases with increasing temperature and with increasing particle loading. The gap conductance increases with increasing temperature and decrease with increasing gap distance. A good gap conductance was observed with helium fill gas. It was seen that the gap conductance was dependent on the thermal conductivity of fill gas and conductance by radiation and could be neglected the conductance through solid-solid contact points of fuel compact and graphite sleeve. (author)

  6. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.; Cramer, G.T.

    1978-06-01

    Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed

  7. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    Holder, N.D.; Lessig, W.S.

    1984-08-01

    GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

  8. Measurement of Weight of Kernels in a Simulated Cylindrical Fuel Compact for HTGR

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Lee, Young Woo; Kim, Young Min; Kim, Yeon Ku; Eom, Sung Ho; Jeong, Kyung Chai; Cho, Moon Sung; Cho, Hyo Jin; Kim, Joo Hee

    2011-01-01

    The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. The weight of fuel kernels in an element is generally measured by the chemical analysis or a gamma-ray spectrometer. Although it is accurate to measure the weight of kernels by the chemical analysis, the samples used in the analysis cannot be put again in the fabrication process. Furthermore, radioactive wastes are generated during the inspection procedure. The gamma-ray spectrometer requires an elaborate reference sample to reduce measurement errors induced from the different geometric shape of test sample from that of reference sample. X-ray computed tomography (CT) is an alternative to measure the weight of kernels in a compact nondestructively. In this study, X-ray CT is applied to measure the weight of kernels in a cylindrical compact containing simulated TRISO-coated particles with ZrO 2 kernels. The volume of kernels as well as the number of kernels in the simulated compact is measured from the 3-D density information. The weight of kernels was calculated from the volume of kernels or the number of kernels. Also, the weight of kernels was measured by extracting the kernels from a compact to review the result of the X-ray CT application

  9. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

    In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

  10. Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

    The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

  11. Quality control procedures for HTGR fuel element components

    International Nuclear Information System (INIS)

    Delle, W.W.; Koizlik, K.; Luhleich, H.; Nickel, H.

    1976-08-01

    The growing use of nuclear reactors for the production of electric power throughout the world, and the consequent increase in the number of nuclear fuel manufacturers, is giving enhanced importance to the consideration of quality assurance in the production of nuclear fuels. The fuel is the place, where the radioactive fission products are produced in the reactor and, therefore, the integrity of the fuel is of utmost importance. The first and most fundamental means of insuring that integrity is through the exercise of properly designed quality assurance programmes during the manufacture of the fuel and other fuel element components. The International Atomic Energy Agency therefore conducted an International Seminar on Nuclear Fuel Quality Assurance in Oslo, Norway from 24 till 28 May, 1976. This KFA report contains a paper which was distributed preliminary during the seminar and - in the second part - the text of the oral presentation. The paper gives a summary of the procedures available in the present state for the production control of HTGR core materials and of the meaning of the particular properties for reactor operation. (orig./UA) [de

  12. Acid in perchloroethylene scrubber solutions used in HTGR fuel preparation processes. Analytical chemistry studies

    International Nuclear Information System (INIS)

    Lee, D.A.

    1979-02-01

    Acids and corrosion products in used perchloroethylene scrubber solutions collected from HTGR fuel preparation processes have been analyzed by several analytical methods to determine the source and possible remedy of the corrosion caused by these solutions. Hydrochloric acid was found to be concentrated on the carbon particles suspended in perchloroethylene. Filtration of carbon from the scrubber solutions removed the acid corrosion source in the process equipment. Corrosion products chemisorbed on the carbon particles were identified. Filtered perchloroethylene from used scrubber solutions contained practically no acid. It is recommended that carbon particles be separated from the scrubber solutions immediately after the scrubbing process to remove the source of acid and that an inhibitor be used to prevent the hydrolysis of perchloroethylene and the formation of acids

  13. Development of processes and equipment for the refabrication of HTGR fuels

    International Nuclear Information System (INIS)

    Sease, J.D.; Lotts, A.L.

    1976-06-01

    Refabrication is in the step in the HTGR thorium fuel cycle that begins with a nitrate solution containing 238 U and culminates in the assembly of this material into fuel elements for use in an HTGR. Refabrication of HTGR fuel is essentially a manufacturing operation and consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and SiC, preparation of fuel rods, and assembly of fuel rods in fuel elements. All the equipment for refabrication of 238 U-containing fuel must be designed for completely remote operation and maintenance in hot cell facilities. This paper describes the status of processes and equipment development for the remote refabrication of HTGR fuels. The feasibility of HTGR refabrication processes has been proven by laboratory development. Engineering-scale development is now being performed on a unit basis on the majority of the major equipment items. Engineering-scale equipment described includes full-scale resin loading equipment, a 5-in.-dia (0.13-m) microsphere coating furnace, a fuel rod forming machine, and a cure-in-place furnace

  14. Thermodynamic assessment of the HTGR fuel system Th-U-C-O

    International Nuclear Information System (INIS)

    Ugajin, M.; Shiba, K.

    1978-01-01

    Carbon monoxide pressures and uranium segregation at 2000 K have been calculated for the three-phase equilibria [(ThU)O 2 + (ThU)C 2 + C] in the Th-U-C-O system. This study is concerned with the thermochemical behavior of (Th, U)O 2 particle fuel for the high-temperature gas-cooled reactor (HTGR). The following two points are considered: (1) Reduction of the in-particle CO pressure of (Th, U)O 2 kernels by doping (Th, U)C 2 to make it an oxygen getter. (2) Prediction of U segregation between (Th, U)O 2 and (Th, U)C 2 , doped in the kernel. (Auth.)

  15. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

  16. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

    International Nuclear Information System (INIS)

    1981-05-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan

  17. HTGR Fuel-Technology Program. Semiannual report for the period ending September 30, 1982

    International Nuclear Information System (INIS)

    1982-11-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at GA Technologies Inc. during the second half of FY-1982. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was completed during this period was the preparation of input data for a long-range technology program plan

  18. Thermal stress analysis of HTGR fuel and control rod fuel blocks in the HTGR in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new approach that utilizes the equivalent solid plate method has been applied to the thermal stress analysis of HTGR fuel and control rod fuel blocks. Cases were considered where these blocks, loaded with reprocessed HTGR fuel pellets, were being cured at temperatures up to 1800 0 C. A two-dimensional segment of a fuel block cross section including fuel, coolant holes, and graphite matrix was analyzed using the ORNL HEATING3 heat transfer code to determine the temperature-dependent effective thermal conductivity for the perforated region of the block. Using this equivalent conductivity to calculate the temperature distributions through different cross sections of the blocks, two-dimensional thermal-stress analyses were performed through application of the equivalent solid plate method. In this approach, the perforated material is replaced by solid homogeneous material of the same external dimensions but whose material properties have been modified to account for the perforations

  19. Criticality considerations for 233U fuels in an HTGR fuel refabrication facility

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-01-01

    Eleven 233 U solution critical assemblies spanning an H/ 233 U ratio range of 40 to 2000 and a bare metal 233 U assembly have been calculated with the ENDF/B-IV and Hansen-Roach cross sections. The results from these calculations are compared with the experimental results and with each other. An increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/ 233 U ratio was observed, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range; whereas the ENDF/B calculations yielded high values for k/sub eff/ on assemblies of low moderation. It is concluded that serious problems exist in the ENDF/B-IV representation of the 233 U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, it is recommended that an experimental program be undertaken to obtain 233 U criticality data at low H/ 233 U ratios for verification of generalized criticality safety guidelines. Part II of this report presents the results of criticality calculations on specific pieces of equipment required for HTGR fuel refabrication. In particular, fuel particle storage hoppers and resin carbonization furnaces are criticality safe up to 22.9 cm (9.0 in.) in diameter providing water or other hydrogenous moderators are excluded. In addition, no criticality problems arise due to accumulation of particles in the off-gas scrubber reservoirs provided reasonable administrative controls are exercised

  20. HTGR fuel rods: carbon-carbon composites designed for high weight and low strength

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1977-01-01

    The evolution of the process for fabricating fuel rods for the high-temperature gas-cooled reactor (HTGR) by injection and carbonization of a thermoplastic matrix that bonds close-packed beds of pyrocarbon-coated fuel particles together is reviewed for the fresh-fuel cycle, and a variant process involving a thermosetting matrix that would allow free-standing carbonization of refabricated fuel is discussed. Previous attempts to fabricate such injection-bonded fuel rods from undiluted thermosetting binders filled with powdered graphite were unsuccessful, because of damage to coatings on fuel particles that resulted from strong particle-to-matrix bonding in conjunction with large matrix shrinkage on carbonization and subsequent irradiation. These problems have now been overcome through the use of a diluted thermosetting matrix with a low-char-yield additive (fugitive), which produces a more porous char similar to that from the pitch-based thermoplastic used in fabrication of fresh fuel. A 1-to-1 dilution of resin with fugitive produced the optimum binder for injection and carbonization, where the fired matrix in such rods contained about 20 wt% binder char and 80 wt% powdered graphite. Thermosetting fuel rods diluted with various amounts of fugitive to give binder chars that range from 12 to 48 wt% of the fired matrix have been subjected to irradiation screening tests, and rods with no more than 32 wt% binder char appear to perform about as well under irradiation as do pitch-based rods. However, particle damage does begin to occur in those lightly diluted rods in which the less-stable binder char constitutes more than 32 wt% of the fired matrix. (author)

  1. Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

    International Nuclear Information System (INIS)

    Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

    1976-02-01

    The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

  2. Design and main characteristics of HTGR fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

    1983-01-01

    Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

  3. Computer simulation of HTGR fuel microspheres using a Monte-Carlo statistical approach

    International Nuclear Information System (INIS)

    Hedrick, C.E.

    1976-01-01

    The concept and computational aspects of a Monte-Carlo statistical approach in relating structure of HTGR fuel microspheres to the uranium content of fuel samples have been verified. Results of the preliminary validation tests and the benefits to be derived from the program are summarized

  4. 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

  5. Performance of HTGR fuel in HFIR capsule HT-33

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Robbins, J.M.

    1979-06-01

    Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO 2 particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375 0 C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400 0 C

  6. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    International Nuclear Information System (INIS)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  7. Irradiation performance of HTGR fuel in HFIR capsule HT-31

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B.; Morgan, C.S.

    1979-05-01

    The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600 0 C, fast neutron fluences (0.18 MeV) up to 9 x 10 25 n/m 2 , and burnups up to 8.9% FIMA for ThO 2 particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO 2 . Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO 2 particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200 0 C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles

  8. Strategy to support HTGR fuel for the 10 MW Indonesia’s experimental power reactor (RDE)

    International Nuclear Information System (INIS)

    Taswanda Taryo; Geni Rina Sunaryo; Ridwan; Meniek Rachmawati

    2018-01-01

    The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power(25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and road map for the preparation of the RDE fuel plant construction with the involvement of national stake holders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia. (author)

  9. Thermal design and analysis of the HTGR fuel element vertical carbonizing and annealing furnace

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1977-06-01

    Computer analyses of the thermal design for the proposed HTGR fuel element vertical carbonizing and annealing furnace were performed to verify its capability and to determine the required power input and distribution. Although the furnace is designed for continuous operation, steady-state temperature distributions were obtained by assuming internal heat generation in the fuel elements to simulate their mass movement. The furnace thermal design, the analysis methods, and the results are discussed herein

  10. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    International Nuclear Information System (INIS)

    Dodd, D.H.; Hienen, J.F.A. van.

    1995-10-01

    This report presents the results of task B.3 of the 'Technology Assessment of the High Temperature Reactor' project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL)

  11. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, D.H.; Hienen, J.F.A. van

    1995-10-01

    This report presents the results of task B.3 of the `Technology Assessment of the High Temperature Reactor` project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL).

  12. Reprocessing yields and material throughput: HTGR recycle demonstration facility

    International Nuclear Information System (INIS)

    Holder, N.; Abraham, L.

    1977-08-01

    Recovery and reuse of residual U-235 and bred U-233 from the HTGR thorium-uranium fuel cycle will contribute significantly to HTGR fuel cycle economics and to uranium resource conservation. The Thorium Utilization National Program Plan for HTGR Fuel Recycle Development includes the demonstration, on a production scale, of reprocessing and refabrication processes in an HTGR Recycle Demonstration Facility (HRDF). This report addresses process yields and material throughput that may be typically expected in the reprocessing of highly enriched uranium fuels in the HRDF. Material flows will serve as guidance in conceptual design of the reprocessing portion of the HRDF. In addition, uranium loss projections, particle breakage limits, and decontamination factor requirements are identified to serve as guidance to the HTGR fuel reprocessing development program

  13. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  14. Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  15. Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy

    International Nuclear Information System (INIS)

    Fukaya, Y.; Goto, M.

    2017-01-01

    Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR

  16. Advanced Fuel UCO Preparation Technology for HTGR (Characteristics of Carbon Black)

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Oh, S. C.; Kim, Y. K.; Cho, M. S.; Kim, W. K.; Kim, Y. M.; Lee, Y. W.; Cho, H. J.; Shin, E. J.

    2010-06-01

    NGNP program for high specification of HTGR nuclear fuel through the GEN IV study is be progressed. Furthermore, because the NGNP program have a highly focused goal like UCO kernel, kernel fabrication and coating types varied which made selection of a US reference fabrication process. In this study, it was evaluated from the reviews on the UO2 and UCO kernel fabrication technologies and its particle characteristics. For improving the UCO qualities, first it was improved the kernel fabrication processes and carbon dispersion method also. New method for carbon dispersion in broth solution was developed, and its characteristics was evaluated from the AGR irradiation tests used the UCO kernel. In fabrication process, also process parameter variation tests in both forming and sintering steps led to an increased understanding of the acceptable ranges for process parameters and additional reduction in required operating times. Another result of this test program was to double the kernel production rate. Following the development tests, approximately 40 kg of natural uranium UCO kernels have been produced for use in coater scale up tests, and approximately 10 kg of low enriched uranium UCO kernels for use in the AGR-2 experiment

  17. The chemical stability of TRISO-coated HTGR fuel. Pt. 1. Status report

    International Nuclear Information System (INIS)

    Groot, P.; Cordfunke, E.H.P.; Konings, R.J.M.

    1994-12-01

    The US fuel seemed to be more difficult to produce than the German fuel. Also the chemical stability of this fuel must be investigated. The conditions are more severe in the US concept than in the German concept. Oxidation of the graphite seems to be no problem, according to US HTGR concept. A ZrC coating seems to have a number of advantages with regard to the SiC coating: (1) Better retention, (2) no reaction with Pd, (3) no thermal dissociation. Only the oxidation resistance is worse than SiC. Also the maximum stress must be determined that the ZrC coating can have. (orig./HP)

  18. Experimental determination of the Koo fuel temperature coefficient for an HTGR lattice

    Energy Technology Data Exchange (ETDEWEB)

    Agostini, P.; Benedetti, F.; Brighenti, G.; Chiodi, P. L.; Dell' Oro, P.; Giuliani, C.; Tassan, S.

    1974-10-15

    This paper describes temperature-dependent k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  19. 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.

  20. 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

  1. Uranium loss from BISO-coated weak-acid-resin HTGR fuel

    International Nuclear Information System (INIS)

    Pearson, R.L.; Lindemer, T.B.

    1977-02-01

    Recycle fuel for the High-Temperature Gas-Cooled Reactor (HTGR) contains a weak-acid-resin (WAR) kernel, which consists of a mixture of UC 2 , UO 2 , and free carbon. At 1900 0 C, BISO-coated WAR UC 2 or UC 2 -UO 2 kernels lose a significant portion of their uranium in several hundred hours. The UC 2 decomposes and uranium diffuses through the pyrolytic coating. The rate of escape of the uranium is dependent on the temperature and the surface area of the UC 2 , but not on a temperature gradient. The apparent activation energy for uranium loss, ΔH, is approximately 90 kcal/mole. Calculations indicate that uranium loss from the kernel would be insignificant under conditions to be expected in an HTGR

  2. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  3. Study on reprocessing of uranium-thorium fuel with solvent extraction for HTGR

    International Nuclear Information System (INIS)

    Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun

    1992-08-01

    A single cycle process by solvent extraction with acid feed solution is suggested. The purpose is to reprocess uranium-thorium fuel elements which are of high burn-up and rich of 232 U from HTGR (high temperature gas cooled reactor). The extraction cascade tests have been completed. The recovery of uranium and thorium is greater than 99.6%. By this method, the requirement, under remote control to re-fabricate fuel elements, of decontamination factors for Cs, Sr, Zr-Nb and Ru has been reached

  4. Process behavior and environmental assessment of 14C releases from an HTGR fuel reprocessing facility

    International Nuclear Information System (INIS)

    Snider, J.W.; Kaye, S.V.

    1976-01-01

    Large quantities of 14 CO 2 will be evolved when graphite fuel blocks are burned during reprocessing of spent fuel from HTGR reactors. The possible release of some or all of this 14 C to the environment is a matter of concern which is investigated in this paper. Various alternatives are considered in this study for decontaminating and releasing the process off-gas to the environment. Concomitant radiological analyses have been done for the waste process scenarios to supply the necessary feedbacks for process design

  5. TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

    1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

  6. Scoping study of flowpath of simulated fission products during secondary burning of crushed HTGR fuel in a quartz fluidized-bed burner

    International Nuclear Information System (INIS)

    Rindfleisch, J.A.; Barnes, V.H.

    1976-04-01

    The results of four experimental runs in which isotopic tracers were used to simulate fission products during fluidized bed secondary burning of HTGR fuel were studied. The experimental tests provided insight relative to the flow path of fission products during fluidized-bed burning of HTGR fuel

  7. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  8. HTGR fuel development: investigations of breakages of uranium-loaded weak acid resin microspheres

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.

    1977-11-01

    During the HTGR fuel development program, a high percentage of uranium-loaded weak acid resin microspheres broke during pneumatic transfer, carbonization, and conversion. One batch had been loaded by the UO 3 method; the other by the ammonia neutralization method. To determine the causes of failure, samples of the two failed batches were investigated by optical microscopy, scanning electron microscopy, electron beam microprobe, and other techniques. Causes of failure are postulated and methods are suggested to prevent recurrence of this kind of failure

  9. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  10. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

  11. Formation of actinides in irradiated HTGR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    dos Santos, A. M.

    1976-03-15

    Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for an actinide isolation were tested with highly irradiated ThO/sub 2/. Separation and decontamination factors are presented. Build-up of /sup 232/U was discussed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal under consideration of the reprocessing technology which is available presently.

  12. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    Verfondern, K.; Martin, R.C.; Moormann, R.

    1993-01-01

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

  13. Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

    International Nuclear Information System (INIS)

    Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

    1977-05-01

    A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

  14. Formation of actinides in irradiated HTGR fuel elements

    International Nuclear Information System (INIS)

    Santos, A.M. dos.

    1976-03-01

    Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for actinide isolation were tested with highly irradiated ThO 2 . Separation and decontamination factors are presented. Actinide nuclide formation can be described by exponential functions of the type ln msub(nuclide) = A + B x % fifa. The empirical factors A and B were calculated performing a least squares analysis. Build-up of 232 U was discussed. According to the experimental results, 232 U is mainly produced from 230 Th, a certain amount (e.g. about 20% at a 10 5 MWd/t burnup) originated from a (n,2n) reaction of 233 U; a formation from 233 Th by a (n,2n) followed by a (n,γ) reaction was not observed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. After a 1,000 years' storage time, the elements Pa, Am and Cm will no longer influence the total hazard index. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal in consideration of the reprocessing technology which is available presently. (orig.) [de

  15. Design and evaluation of an on-line fuel rod assay device for an HTGR fuel refabrication plant

    International Nuclear Information System (INIS)

    Rushton, J.E.; Allen, E.J.; Chiles, M.M.; Jenkins, J.D.

    1979-11-01

    Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233 U and/or 235 U. The fuel rods are approx. 16 mm in diameter and 62 mm long. A typical commercial fuel refabrication facility will have six fuel rod production lines, each producing approximately one fuel rod every 4 seconds at design capacity. One on-line assay device will be present for each two production lines. The relative standard deviation in an individual fuel rod fissile material measurement must be less than 3% to satisfy process and quality control requirements. Systematic errors must be kept less than approx. 0.3% for fissile material measured in fuel rods produced over two months to satisfy material accountability requirements. Several nondestructive assay (NDA) methods were investigated. Because the gamma-ray activity of the refabricated fuel is relatively high due to the presence of 232 U in the fuel and because the gamma-ray activity is not directly related to total or fissile uranium content, NDA methods employing gamma-ray detection did not appear practicable. A method using thermal neutron irradiation and fast-fission neutron detection was selected. An experimental assay device was fabricated based on this NDA method. Experiments were performed to determine the precision and accuracy of the measurements and to investigate potential interferences and systematic errors. Operating procedures were evaluated, and analysis procedures were identified

  16. High-temperature gas reactor (HTGR) market assessment, synthetic fuels analysis

    International Nuclear Information System (INIS)

    1980-08-01

    This study is an update of assessments made in TRW's October 1979 assessment of overall high-temperature gas-cooled reactor (HTGR) markets in the future synfuels industry (1985 to 2020). Three additional synfuels processes were assessed. Revised synfuel production forecasts were used. General environmental impacts were assessed. Additional market barriers, such as labor and materials, were researched. Market share estimates were used to consider the percent of markets applicable to the reference HTGR size plant. Eleven HTGR plants under nominal conditions and two under pessimistic assumptions are estimated for selection by 2020. No new HTGR markets were identified in the three additional synfuels processes studied. This reduction in TRW's earlier estimate is a result of later availability of HTGR's (commercial operation in 2008) and delayed build up in the total synfuels estimated markets. Also, a latest date for HTGR capture of a synfuels market could not be established because total markets continue to grow through 2020. If the nominal HTGR synfuels market is realized, just under one million tons of sulfur dioxide effluents and just over one million tons of nitrous oxide effluents will be avoided by 2020. Major barriers to a large synfuels industry discussed in this study include labor, materials, financing, siting, and licensing. Use of the HTGR intensifies these barriers

  17. Fuel particle coating data

    International Nuclear Information System (INIS)

    Hollabaugh, C.M.; Wagner, P.; Wahman, L.A.; White, R.W.

    1977-01-01

    Development of coating on nuclear fuel particles for the High-Temperature Fuels Technology program at the Los Alamos Scientific Laboratory included process studies for low-density porous and high-density isotropic carbon coats, and for ZrC and ''alloy'' C/ZrC coats. This report documents the data generated by these studies

  18. R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

    1999-09-01

    The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

  19. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  20. ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

    International Nuclear Information System (INIS)

    Fuller, L.C.; Myers, M.L.

    1975-01-01

    1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

  1. Design method of control system for HTGR fuel handling process with control Petri net

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

    As a complex mechanical system,the fuel handling system (FHS) of pebble-bed high temperature gas cooled reactor (HTGR) is with the features of complicated structure, numerous control devices and strict working scheduling. It is very important to precisely describe the function of FHS and effectively design its control system. A design method of control system based on control Petri net (CPN) is introduced in this paper. By associating outputs and operations with places, associating inputs and conditions with transitions, and introducing macro-places and macro-actions, the CPN realizes hierarchy design of complex control system. Based on the analysis of basic functions and working flow of FHS, its control system is described and designed by CPN. According to the firing regulation of transition,the designed CPN can be easily converted into LAD program of PLC, which can be implemented on the FHS simulating control test-bed. Application illuminates that proposed method has the advantages of clear design structure, exact description power and effective design ability of control program, which can meet the requirements of FHS control sys-tem design. (authors)

  2. Particle fuel bed tests

    International Nuclear Information System (INIS)

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

    1985-01-01

    Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

  3. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  4. Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

    International Nuclear Information System (INIS)

    Kim, Young Min; Lee, Y. W.; Chang, J. H.

    2005-01-01

    IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

  5. Comparison of HTGR fuel design, manufacture and quality control methods between Japan and China

    International Nuclear Information System (INIS)

    Fu Xioming; Takahashi, Masashi; Ueta, Shouhei; Sawa, Kazuhiro

    2002-05-01

    The first-loading fuel for the HTTR was started to fabricate at Nuclear Fuel Industries (NFI) in 1995 and the HTTR reached criticality in 1998. Meanwhile, 10 MW high temperature reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, and the first-loading fuel was fabricated concurrently. The HTR-10 reached criticality in December 2000. Though fuel type is different, i.e., pin-in-block type for the HTTR and pebble bed type for the HTR-10, the fabrication method of TRISO coated fuel particles is similar to each other. This report describes comparison of fuel design, fabrication process and quality inspection between them. (author)

  6. Development of coated particle fuel technology

    International Nuclear Information System (INIS)

    Cho, Moonsung; Kim, B. G.; Kim, D. J.

    2011-06-01

    Ammonia contacting method for prehardenning the surfaces of ADU liquid droplets and the ageing/washing/drying method and equipment for spherical dried-ADU particles were improved and tested with laboratory sacle. After the improvement of fabrication process, the sphericity of UO 2 kernel obtained to 1.1, and the sintered density and O/U ratio of final UO 2 kernel were above 10.60g/cm 3 . 2.01 respectively. Defects of SiC coating layer could be minimized by optimization of gas flow rate. The fracture strength of SiC layer increased from 450 MPa to 530 MPa by controlling the coating defects. An effort was made to develop the fundamental technology for the fuel element compact for use in High Temperature Gas-cooled Reactor(HTGR) through an establishment of fabrication process, required materials and process equipment as well as performing experiments to identify the basic process conditions and optimize them. Thermal load simulation and verification experiments were carried out for an assesment of the design feasibility of the irradiation rod. Out-of-pile testing of irradiation device such as measurement of pressure drop and vibration, endurance test was performed and the validity of its design was confirmed. A fuel performance analysis code, COPA has been developed to calculate the fuel temperature, the failure fractions of coated fuel particles, the release of fission products. The COPA code can be used to evaluate the performance of the high temperature reactor fuel under the reactor operation, irradiation, heating conditions. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. QC technology was established for TRISO-coated fuel particle. A method for accurate measurement of the optical anisotropy factor for PyC layers of coated particles was developed. Technology and inspection procedures for density

  7. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    International Nuclear Information System (INIS)

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

  8. FRESCO-II: A computer program for analysis of fission product release from spherical HTGR-fuel elements in irradiation and annealing experiments

    International Nuclear Information System (INIS)

    Krohn, H.; Finken, R.

    1983-06-01

    The modular computer code FRESCO has been developed to describe the mechanism of fission product release from a HTGR-Core under accident conditions. By changing some program modules it has been extended to take into account the transport phenomena (i.e. recoil) too, which only occur under reactor operating conditions and during the irradiation experiments. For this report, the release of cesium and strontium from three HTGR-fuel elements has been evaluated and compared with the experimental data. The results show that the measured release can be described by the considered models. (orig.) [de

  9. Performance of HTGR fertile particles irradiated in HFIR capsule HT-32

    International Nuclear Information System (INIS)

    Long, E.L. Jr.; Robbins, J.M.; Tiegs, T.N.; Kania, M.J.

    1980-04-01

    The HT-32 experiment was an uninstrumented capsule irradiated for four cycles in the target position of the High-Flux Isotope Reactor (HFIR). The experiment was designed to: provide supplemental simulated fuel rods for thermal transport and expansion measurements; test fertile kernels with Al 2 O 3 and SiO 2 additives for improved fission product retention; study the stability and permeability of low-temperature isotropic (LTI) pyrocarbon coatings; test Biso- and Triso-coatings derived in a large (0.24-m-dia) coating furnace with a frit distributor; investigate the performance of particles with an outer layer of SiC both as loose particles and as resin-bonded fuel rods; and evaluate high-density alumina as a potential high-temperature thermometry sheathing material

  10. Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)

  11. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  12. Burning of spent fuel of an accelerator-driven modular HTGR in sub-critical condition

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Chang Hong; Wu Zongxin; Gu Yuxiang

    2002-01-01

    The modular high temperature gas cooled reactor (MHTGR) has good safety characteristics because of the use of coated particles in the fuel element. After the particles cool outside of the reactor for some time, the spent fuel can be re-utilized. The author describes a physics feasibility study for the burning of spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor in an accelerator-driven sub-critical reactor. A conceptual design is given for the 30 MW accelerator-driven sub-critical reactor. The neutron transport in the sub-critical reactor was simulated using the MCNP code, and the burnup was calculated using the ORIGEN2 code. The results show that the accelerator-driven sub-critical gas-cooled reactor has reliable sub-criticality and low power density and that the spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor can be burned to provide 20% more energy

  13. Waste management considerations in HTGR recycle operations

    International Nuclear Information System (INIS)

    Pence, D.T.; Shefcik, J.J.; Heath, C.A.

    1975-01-01

    Waste management considerations in the recycle of HTGR fuel are different from those encountered in the recycle of LWR fuel. The types of waste associated with HTGR recycle operations are discussed, and treatment methods for some of the wastes are described

  14. Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Thoms, K.R.

    1979-03-01

    Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimens with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC 2

  15. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, B. G.; Kim, S. H.

    2007-06-01

    Uranium kernel fabrication technology using a wet chemical so-gel method, a key technology in the coated particle fuel area, is established up to the calcination step and the first sintering of UO2 kernel was attempted. Experiments on the parametric study of the coating process using the surrogate ZrO2 kernel give the optimum conditions for the PyC and SiC coating layer and ZrC coating conditions were obtained for the vaporization of the ZrCl4 precursor and coating condition from ZrC coating experiments using plate-type graphite substrate. In addition, by development of fuel performance analysis code a part of the code system is completed which enables the participation to the benchmark calculation and comparison in the IAEA collaborated research program. The technologies for irradiation and post irradiation examination, which are important in developing the HTGR fuel technology of its first kind in Korea was started to develop and, through a feasibility study and preliminary analysis, the technologies required to be developed are identified for further development as well as the QC-related basic technologies are reviewed, analyzed and identified for the own technology development. Development of kernel fabrication technology can be enhanced for the remaining sintering technology and completed based on the technologies developed in this phase. In the coating technology, the optimum conditions obtained using a surrogate ZrO2 kernel material can be applied for the uranium kernel coating process development. Also, after completion of the code development in the next phase, more extended participation to the international collaboration for benchmark calculation can be anticipated which will enable an improvement of the whole code system. Technology development started in this phase will be more extended and further focused on the detailed technology development to be required for the related technology establishment

  16. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  17. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bohnenstingl, J.; Djoa, S. H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-04-15

    This paper describes a process developed for the retainment and separation of volatile (3H, 129 +131I) and gaseous (85Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 deg K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 deg K and low subpressure; deposition of krypton in solid form at 80 deg K after compression to about 6 bar; decontamination of 85krypton-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, e.g., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1/3 of the full capacity and can treat about 1 m3 STP/h helium, corresponding to a quantity of about 10,000 MW(e) HTGR-fuel reprocessing plant.

  18. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    International Nuclear Information System (INIS)

    Bohnenstingl, J.; Djoa, S.H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-01-01

    This paper describes a process developed for the retainment and separation of volatile ( 3 H, 129+131 I) and gaseous ( 85 Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 0 K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 0 K and low subpressure; deposition of krypton in solid form at 80 0 K after compression to about 6 bar; decontamination of 85 Kr-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, i.e., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1 / 3 of the full capacity and can treat about 1 m 3 STP/h helium, corresponding to a quantity of about 10,000 MW/sub e/ HTGR-fuel reprocessing plant

  19. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

    The reference fuel kernel for recycle of 233 U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233 UO 2 (NO 3 ) 2 solution from a fuel reprocessing plant contains excess HNO 3 (NO 3 - /U ratio of approximately 2.2). The reference flowsheet for a 233 U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO 3 - /U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  20. A study on coated particle fuel properties and performances and phase-I data base establishment

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Hyo Cheol; Im, Byeong Ju; Yun, Sang Pil; Son, Seung Beom; Lee, Gyeong Hui; Jang, Jeong Nam

    2006-03-01

    For the successful development of the high temperature gas cooled reactor acquisition and generation of the high temperature properties of reactor materials, especially temperature and burn-up dependent properties of coated particle fuel and fuel element, are crucially essential. Recently national project for HTGR for hydrogen production has been kicked off. However, we have had little experience on this new challenges. Therefore, it became necessary to build up the materials properties and fuel performance data base. In this study, a primitive properties and performance DB for coated particle fuel was developed. This database report consists two sections: materials properties and fuel performance. The materials properties has three parts: kernel materials, carbide coating materials, and fuel elements and graphite matrix. UO 2 and UCO belong to kernel materials while PyC, SiC, and ZrC comprises the coating materials section. Thermal, mechanical and physical properties of these materials were collected, reviewed, and summarized. Additionally, the property change induced by manufacture process and irradiation were collected and summarized. Performance data were also collected, reviewed, and analyzed based on the key phenomena and failure mechanism. All of these data will be accessible in the on-line system. These results will be directly used for HTGR fuel design and fabrication and preliminary fuel performance analysis under irradiation

  1. HTGR safety research program

    International Nuclear Information System (INIS)

    Barsell, A.W.; Olsen, B.E.; Silady, F.A.

    1981-01-01

    An HTGR safety research program is being performed supporting and guided in priorities by the AIPA Probabilistic Risk Study. Analytical and experimental studies have been conducted in four general areas where modeling or data assumptions contribute to large uncertainties in the consequence assessments and thus, in the risk assessment for key core heat-up accident scenarios. Experimental data have been obtained on time-dependent release of fission products from the fuel particles, and plateout characteristics of condensible fission products in the primary circuit. Potential failure modes of primarily top head PCRV components as well as concrete degradation processes have been analyzed using a series of newly developed models and interlinked computer programs. Containment phenomena, including fission product deposition and potential flammability of liberated combustible gases have been studied analytically. Lastly, the behaviour of boron control material in the core and reactor subcriticality during core heatup have been examined analytically. Research in these areas has formed the basis for consequence updates in GA-A15000. Systematic derivation of future safety research priorities is also discussed. (author)

  2. Development of a surveillance robot for dimensional and visual inspection of fuel and reflector elements from the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Marsh, N.I.; Miller, C.M.; Saurwein, J.J.; Smith, T.L.

    1979-11-01

    A robotic device has been developed for dimensional and visual inspection of irradiated HTGR core components. The robot consists of a rotary table and a two-finger probe, driven by stepping motors, and four remotely controlled television cameras. Automated operation is accomplished via minicomputer control. A total of 51 irradiated fuel and reflector elements were inspected at a fraction of the time and cost required for conventional methods

  3. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength for the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

  4. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance and fuel particle chemical performande. (orig.) [de

  5. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance

  6. Conceptual design of the special nuclear material nondestructive assay and accountability system for the HTGR fuel refabrication pilot plant

    International Nuclear Information System (INIS)

    Jenkins, J.D.; McNeany, S.R.; Rushton, J.E.

    1975-07-01

    The conceptual design of the fissile material assay and accountability system for the HTGR refabrication pilot plant has been established. The primary feature affecting the design is the high, time varying, gamma activity of the process material due to the unavoidable presence of uranium-232. This imposes stringent requirements for remote operation and remote maintainability of system components. At the same time, the remote operation lends itself to implementation of an automated data collection and processing system for real-time accountability. The high time-varying gamma activity of the material also precludes application of a number of techniques presently employed for light-water reactor fuel assay. The techniques selected for application in the refabrication facility are (1) active thermal neutron interrogation with fast-fission or delayed-neutron counting for fuel-rod and small-sample assay, (2) calorimetry for high-level waste assay, and (3) passive gamma scanning for low-level waste assay, and rapid on-line relative rod-loading measurements. The principal nondestructive assay subsystems are identified as (1) on-line devices for 100 percent product fuel rod assay and quality control, (2) a multipurpose device in the sample inspection laboratory for small- sample assay and secondary standards calibration, and (3) equipment for assay of high- and low-uranium content scrap and waste materials. A data processing system, which coordinates data from these subsystems with information from other process control sensors, is included to provide real-time material balance information. (U.S.)

  7. A study on properties-performances of coated particle fuel and on-line DB establishment

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Hyo Cheol; Jang, Jeong Nam; Kwon, Seok Hwan

    2007-03-01

    Recently national project for HTGR for hydrogen production has been kicked off. However, For the successful development of the high temperature gas cooled reactor high temperature and burn-up dependent properties of the reactor materials are essentially and crucially required. Therefore, it was proposed to build up the materials properties and fuel performance data base. In this study, a phase - 1 properties and performance DB for coated particle fuel was developed. This database report consists two sections: materials properties and fuel performance. The materials properties has three parts: kernel materials, carbide coating materials, and fuel elements and graphite matrix. UO2 and UCO belong to kernel materials while PyC, SiC, and ZrC comprises the coating materials section. Thermal, mechanical and physical properties data of these materials were collected, reviewed, and summarized. Additionally, the property change induced by manufacture process and irradiation were reviewed. Fuel performance data were also collected, reviewed, and analyzed based on the key phenomena and failure mechanism, These performance data are divided into two: normal and accident. All of these data will be accessible in the pc based stand-alone system. These results will be directly used for HTGR fuel design and fabrication and preliminary fuel performance analysis under irradiation

  8. A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

    International Nuclear Information System (INIS)

    Talamo, A.

    2009-01-01

    In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

  9. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  10. Transfer of fissile material through shielding coatings in emergency heating of HTGR coated particles

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Zhuravkov, S.G.; Koptev, M.A.; Kurepin, A.D.

    1990-01-01

    The measurement results of leakage dynamics of fissile material from the coated particles within a temperature range of 1200 + 2000 deg. C are given. The methods of carrying out the experiments are briefly described. The relation of the leakage rate of uranium-235 from CP (coated particles) with the pyrocarbonic coatings has been obtained. (author)

  11. Postirradiation examination of Peach Bottom HTGR Driver Fuel Element E06-01

    International Nuclear Information System (INIS)

    Dyer, F.F.; Wichner, R.P.; Martin, W.J.; Fairchild, L.L.; Kedl, R.J.; de Nordwall, H.J.

    1976-04-01

    The report presented describes the postirradiation examinations of driver fuel element E06-01, which had been irradiated an equivalent of 384 full-power days in Peach Bottom, Unit 1. The fuel element is described in detail and its temperature and irradiation service history briefly outlined. Results presented include: (1) visual observations; (2) critical dimensions of fuel compacts, sleeve, and spine; (3) axial distributions of gamma-emitting nuclides plus 3 H and 90 Sr; (4) radial distributions of these nuclides in the sleeve and spine at three axial locations in the fueled regions and three locations in the upper reflector; (5) metallographic examination of samples of fuel compact material; and (6) burnup determinations via radiochemical analyses at two compact locations

  12. 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.

  13. HTGR R and D programs

    International Nuclear Information System (INIS)

    Neylan, A.J.; Brisbois, J.

    1979-01-01

    A significant R and D program (including in certain cases full-scale prototype tests) formed the basis for the design and key elements in the foregoing projects and is continuing to provide a basis for generic design development. HTGR R and D programs are both privately and government sponsored. This paper provides an overview of the background, current status and outstanding design issues/problems remaining in the area of NSS Plant, Materials and Fuel. The specific objectives and scope of all recently completed, ongoing and planned major HTGR R and D programs are presented

  14. Use of spikants in HTGR fuel and their effect on costs

    International Nuclear Information System (INIS)

    Brooks, L.H.

    1979-05-01

    The costs of fresh fuel fabrication have been estimated for flowsheets that have spikant added (Co-60) at various points. The costs are compared with refabricated and fresh fuel fabrication costs. It is shown that the costs increase as the spikant is added nearer to the plant feed point. The least cost is achieved by adding a detachable spikant to the finished fuel element. The highest cost is incurred when the spikant is fed in with the plant feed. This cost is greater than that for refabricated fuel because extra process cells are necessary to prepare the spikant and, in addition, the gamma flux from the Co-60 is much greater than from U-232 and greater precautions have to be taken

  15. Distribution of fission products in Peach Bottom HTGR fuel element E01-01

    International Nuclear Information System (INIS)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

    1978-10-01

    The fifth in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is described. The element analyzed received an equivalent of 897 full-power days of irradiation prior to the scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a 137 Cs inventory of 20.3 Ci in the graphite sleeve and 8.1 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides 134 Cs, /sup 110 m/Ag, 60 Co, and 154 Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters 3 H, 14 C, and 90 Sr were obtained at four axial locations of the fueled region of the element sleeve and two axial locations of the element spine. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed

  16. Irradiation performance of HTGR Biso fertile particles in HFIR experiments HT-17, -18, and -19

    International Nuclear Information System (INIS)

    Long, E.L. Jr.; Beatty, R.L.; Robbins, J.M.; Kania, M.J.; Eatherly, W.P.

    1978-11-01

    A series of Biso-coated fertile particles was irradiated in the target facility of the High-Flux Isotope Reactor. The primary objectives of this experiment were to relate the fast-neutron stability of dense propylene-derived pyrocarbons to preferred orientation and to relate irradiation performance to preirradiation characterization values. Coating characterization included x-ray BAF, optical anisotropy, gaseous permeability, small-angle x-ray scattering, and thickness and density determinations. Other objectives were to test Biso-coated large-diameter ThO 2 kernels and coatings derived from propylene diluted with CO 2 rather than argon. Visual examination of the irradiated particles showed that the majority had failed as a result of fast-neutron damage

  17. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  18. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  19. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  20. Coated particle fuel for high temperature gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

    2007-10-15

    Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be{exclamation_point} It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where 9 x 10{sup -4} initial free heavy metal fraction was typical for early AVR carbide fuel and 3 x 10{sup -4} initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/traditional and new materials, manufacturing technologies/ quality control/ quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with 700-750 .deg. C helium coolant gas exit, for gas turbine

  1. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

    2007-01-01

    Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where 9 x 10 -4 initial free heavy metal fraction was typical for early AVR carbide fuel and 3 x 10 -4 initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/traditional and new materials, manufacturing technologies/ quality control/ quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with 700-750 .deg. C helium coolant gas exit, for gas turbine applications at 850-900 .deg. C

  2. 1976 scientific progress report. [Fuel and coating materials for HTGR]; Wissenschaftlicher Ergebnisberict 1976

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, H.

    1976-07-01

    Activities at the Institute for Reactor Materials in the production and properties of high temperature gas cooled reactor fuel and coating materials are summarized. Major emphasis was placed on investigations of pyrocarbon, BISO and TRISO coatings, uranium and thorium oxides and carbides, and graphite and matrix materials. A list of publications is included. (HDR)

  3. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    International Nuclear Information System (INIS)

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  4. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  5. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  6. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

    2016-01-01

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  7. Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

    International Nuclear Information System (INIS)

    Yoshimuda, Hideharu.

    1990-01-01

    Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

  8. SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W.

    2017-07-01

    High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must have very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the

  9. Fuel cycle cost comparison of choices in U-235 recycle in the HTGR

    International Nuclear Information System (INIS)

    Rothstein, M.P.

    1976-07-01

    An analysis of alternative options for the recycle of discharged makeup U-235 (''residual'' makeup) in HTGRs shows that the three-particle system which has been the reference plan remains optimal. This result considers both the resource utilization and the handling costs attendant to the alternative strategies (primarily in the recycle facility and in waste disposal). Furthermore, this result appears to be true under all forseeable economic conditions. A simple risk assessment indicates that recycle cost (including reprocessing, refabrication, and related waste disposal) would have to double or triple in order for the alternative U-235 recycle schemes to become attractive. This induces some degree of confidence in the choice of staying with the reference cycle in spite of the large degree of uncertainty over recycle and its costs

  10. Cesium transport data for HTGR systems

    International Nuclear Information System (INIS)

    Myers, B.F.; Bell, W.E.

    1979-09-01

    Cesium transport data on the release of cesium from HTGR fuel elements are reviewed and discussed. The data available through 1976 are treated. Equations, parameters, and associated variances describing the data are presented. The equations and parameters are in forms suitable for use in computer codes used to calculate the release of metallic fission products from HTGR fuel elements into the primary circuit. The data cover the following processes: (1) diffusion of cesium in fuel kernels and pyrocarbon, (2) sorption of cesium on fuel rod matrix material and on graphite, and (3) migration of cesium in graphite. The data are being confirmed and extended through work in progress

  11. Utilization of Plutonium and Higher Actinides in the HTGR as Possibility to Maintain Long-Term Operation on One Fuel Loading

    International Nuclear Information System (INIS)

    Tsvetkova, Galina V.; Peddicord, Kenneth L.

    2002-01-01

    Promising existing nuclear reactor concepts together with new ideas are being discussed worldwide. Many new studies are underway in order to identify prototypes that will be analyzed and developed further as systems for Generation IV. The focus is on designs demonstrating full inherent safety, competitive economics and proliferation resistance. The work discussed here is centered on a modularized small-size High Temperature Gas-cooled Reactor (HTGR) concept. This paper discusses the possibility of maintaining long-term operation on one fuel loading through utilization of plutonium and higher actinides in the small-size pebble-bed reactor (PBR). Acknowledging the well-known flexibility of the PBR design with respect to fuel composition, the principal limitations of the long-term burning of plutonium and higher actinides are considered. The technological challenges and further research are outlined. The results allow the identification of physical features of the PBR that significantly influence flexibility of the design and its applications. (authors)

  12. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-79 are reported. The report covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop an MEU fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  13. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant.

  14. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    International Nuclear Information System (INIS)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  15. Determination of reactivity of multiplying systems filled with spherical HTGR-fuel elements using kinetic methods with regard to the pulsed-neutron method

    International Nuclear Information System (INIS)

    Drueke, V.

    1978-06-01

    At three critical or subcritical facilities - two of them filled with spherical HTGR-fuel elements - the reactivity is determined using kinetic methods. Besides the inverskinetic method the applicability of the pulsed-neutron method is investigated. The experimental results using the pulsed-neutron method are compared partly with the inverskinetic method and partly with diffusion-calculations. It is shown, that in the HTGR the space dependence of the reactivity in radial direction is not remarkable in spite of the 'kinetic distortion'; on the contrary in axial direction - the direction of the external neutron source - space dependent reactivity worths are measured. The results of the pulsed-neutron methods of Sjoestrand and Simmons-King are rather good applicable in all configurations. For the method of Sjoestrand it is necessary to select the detector positions, whereas for Simmons-King the calculated life-time determines the results. Therefore it is proposed to compare calculated and measured decay constants of the prompt neutron field in future. (orig.) [de

  16. High-temperature Gas Reactor (HTGR)

    Science.gov (United States)

    Abedi, Sajad

    2011-05-01

    General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

  17. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new method for performing thermal stress analyses in structures with multiple penetrations was applied to these analyses. This method couples the development of an equivalent thermal conductivity for the blocks, a technique that has been used extensively for modeling the thermal characteristics of reactor cores, with the use of the equivalent solid plate method for stress analysis. Using this equivalent thermal conductivity, which models as one material the heat transfer characteristics of the fuel, coolant, and graphite two-dimensional, steady-state thermal analyses of the fuel and control rod fuel blocks were performed to establish all temperature boundaries required for the stress analyses. In applying the equivalent solid plate method, the region of penetrations being modeled was replaced by a pseudo material having the same dimensions but whose materials properties were adjusted to account for the penetration. The peak stresses and strains were determined by applying stress and strain intensification factors to the calculated distributions. The condition studied was where the blocks were located near the center of the furnace. In this position, the axial surface of the block is heated near one end and cooled near the other. The approximate axial surface temperatures ranged from 1521 0 C at both the heated and the cooled ends to a peak of 1800 0 C near the center. Five specific cases were analyzed: plane (two-dimensional thermal, plane stress strain) analyses of each end of a standard fuel block (2 cases), plane analyses of each end of a control rod fuel block (2 cases), and a two-dimensional analysis of a fuel block treated as an axisymmetric cylind

  18. Quality control of coated fuel particles for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kaneko, Mitsunobu

    1987-01-01

    The quality control of the coated fuel particles for high temperature gas-cooled reactors is characterized by the fact that the size of the target product to be controlled is very small, and the quantity is very large. Accordingly, the sampling plan and the method of evaluating the population through satisfically treating the measured data of the samples are the important subjects to see and evaluate the quality of a batch or a lot. This paper shows the fabrication process and the quality control procedure for the coated fuel particles. The development work of a HTGR was started by Japan Atomic Energy Research Institute in 1969, and as for the production technology for coated fuel particles, Nuclear Fuel Industries, Ltd. has continued the development work. The pilot plan with the capacity of about 40 kg/year was established in 1972. The fuel product fabricated in this plant was put to the irradiation experiment and out-of-pile evaluation test. In 1983, the production capacity was expanded to 200 kg/year, and the fuel compacts for the VHTRC in JAERI were produced for two years. The basic fuel design, the fabrication process, the quality control, the process control and the quality assurance are reported. For the commercial product, the studies from the viewpoint of production and quality control costs are required. (Kako, I.)

  19. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    Science.gov (United States)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

  20. HTGR generic technology program. Semiannual report ending March 31, 1980

    International Nuclear Information System (INIS)

    1980-05-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an MEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbine and process heat plants

  1. Carbon fuel particles used in direct carbon conversion fuel cells

    Science.gov (United States)

    Cooper, John F.; Cherepy, Nerine

    2012-10-09

    A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

  2. Carbon Fuel Particles Used in Direct Carbon Conversion Fuel Cells

    Science.gov (United States)

    Cooper, John F.; Cherepy, Nerine

    2008-10-21

    A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

  3. Particle Count Limits Recommendation for Aviation Fuel

    Science.gov (United States)

    2015-10-05

    Particle Counter Methodology • Particle counts are taken utilizing calibration methodologies and standardized cleanliness code ratings – ISO 11171 – ISO...Limits Receipt Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18/16/13 14/10/7 Pamas / Parker / Particle Solutions 19/17...12 U.S. DOD 19/17/14/13* Diesel Fuel World Wide Fuel Charter 5th 18/16/13 DEF (AUST) 5695B 18/16/13 Caterpillar 18/16/13 Detroit Diesel 18/16/13 MTU

  4. Characterisation of TRISO fuel particles

    International Nuclear Information System (INIS)

    Lopez H, E.; Yang, D.

    2012-10-01

    The TRISO (tri structural isotropic) coated fuel particle is a key component contributing to the inherent safety of the High Temperature Reactor. A uranium kernel is coated with three layers of pyrolytic carbon and one of silicon carbide. The purpose of these coatings is to work as a miniature fission product containment vessel capable of enclosing all important radio nuclei under normal and off-normal reactor operating conditions. Due to the importance of these coatings, is of great interest to establish characterisation techniques capable of providing a detailed description of their microstructure and physical properties. Here we describe the use of Raman spectroscopy and two modulator generalised ellipsometry to study the anisotropy and thermal conductivity of pyrolytic carbon coatings, as well as the stoichiometry of the silicon carbide coatings and fibres. (Author)

  5. Evolution of Particle Bed Reactor Fuel

    Science.gov (United States)

    Jensen, Russell R.; Evans, Robert S.; Husser, Dewayne L.; Kerr, John M.

    1994-07-01

    To realize the potential performance advantages inherent in a particle bed reactor (PBR) for nuclear thermal propulsion (NTP) applications, high performance particle fuel is required. This fuel must operate safely and without failure at high temperature in high pressure, flowing hydrogen propellant. The mixed mean outlet temperature of the propellant is an important characteristic of PBR performance. This temperature is also a critical parameter for fuel particle design because it dictates the required maximum fuel operating temperature. In this paper, the evolution in PBR fuel form to achieve higher operating temperatures is discussed and the potential thermal performance of the different fuel types is evaluated. It is shown that the optimum fuel type for operation under the demanding conditions in a PBR is a coated, solid carbide particle.

  6. Development, design, and preliminary operation of a resin-feed processing facility for resin-based HTGR fuels

    International Nuclear Information System (INIS)

    Haas, P.A.; Drago, J.P.; Million, D.L.; Spence, R.D.

    1978-01-01

    Fuel kernels for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared by loading carboxylic acid cation exchange resins with uranium and carbonizing at controlled conditions. Resin-feed processing was developed and a facility was designed, installed, and operated to control the kernel size, shape, and composition by processing the resin before adding uranium. The starting materials are commercial cation exchange resins in the sodium form. The size separations are made by vibratory screening of resin slurries in water. After drying in a fluidized bed, the nonspherical particles are separated from spherical particles on vibratory plates of special design. The sized, shape-separated spheres are then rewetted and converted to the hydrogen form. The processing capacity of the equipment tested is equivalent to about 1 kg of uranium per hour and could meet commercial recycle plant requirements without scale-up of the principal process components

  7. HTGR Measurements and Instrumentation Systems

    International Nuclear Information System (INIS)

    Ball, Sydney J.; Holcomb, David Eugene; Cetiner, Mustafa Sacit

    2012-01-01

    This report provides an integrated overview of measurements and instrumentation for near-term future high-temperature gas-cooled reactors (HTGRs). Instrumentation technology has undergone revolutionary improvements since the last HTGR was constructed in the United States. This report briefly describes the measurement and communications needs of HTGRs for normal operations, maintenance and inspection, fuel fabrication, and accident response. The report includes a description of modern communications technologies and also provides a potential instrumentation communications architecture designed for deployment at an HTGR. A principal focus for the report is describing new and emerging measurement technologies with high potential to improve operations, maintenance, and accident response for the next generation of HTGRs, known as modular HTGRs, which are designed with passive safety features. Special focus is devoted toward describing the failure modes of the measurement technologies and assessing the technology maturity.

  8. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  9. GCRA perspective on the HTGR-GT plant configuration

    International Nuclear Information System (INIS)

    1979-06-01

    Design specifications for the HTGR type reactor and gas turbine combination are presented concerning the turbomachinery; generator and isophase bus duct; PCRV and internals; heat exchangers; operability; maintenance; safety and licensing; core design; and fuel design

  10. Detection of gas-permeable fuel particles for highl 7490 temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thiele, B.A.; Stinton, D.P.; Costanzo, D.A.

    1980-01-01

    Fuel for High-Temperature Gas-Cooled Reactors (HTGR) consists of uranium oxide-carbide and thoria microspheres coated with layers of pyrolytic carbon and silicon carbide. The pyrolytic carbon coatings must be gas-tight to perform properly during irradiation. Therefore, particles must be carefully characterized to determine the number of defective particles (ie bare kernels, and cracked or permeable coatings). Although techniques are available to determine the number of bare kernels or cracked coatings, no reliable technique has been available to measure coating permeability. This work describes a technique recently developed to determine whether coatings for a batch of particles are gas-tight or permeable. Although most of this study was performed on Biso-coated particles, the technique applies equally well to Triso-coated particles. About 150 randomly selected Biso-particle batches were studied in this work. These batches were first subjected to an 18-hr chlorination at 15000C, and the volatile thorium tetrachloride released through cracked or very permeable coatings was measured versus chlorination time. Chlorinated batches were also radiographed to detect any thorium that had migrated from the kernel into the coatings. From this work a technique was developed to determine coating permeability. This consists of an 18-hr chlorination of multiple samples without measurement of the heavy metal released. Each batch is then radiographed and the heavy metal diffusion within each particle is examined so it can be determined if a particle batch is permeable, slightly permeable, or gas-tight. (author)

  11. HTGR generic technology program plan (FY 80)

    International Nuclear Information System (INIS)

    1980-01-01

    Purpose of the program is to develop base technology and to perform design and development common to the HTGR Steam Cycle, Gas Turbine, and Process Heat Plants. The generic technology program breaks into the base technology, generic component, pebble-bed study, technology transfer, and fresh fuel programs

  12. 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.

  13. 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.

  14. Interim design report: fuel particle crushing

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

    1977-11-01

    The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

  15. Nuclear fuel particle and method of production

    International Nuclear Information System (INIS)

    Wagner-Loffler, M.

    1975-01-01

    The core consisting of fuel oxide (UO 2 or Th or Pu oxide) of a fuel particle coated with carbon-contained material is enriched with a small addition (max 6 wt.%) of a Ba or Sr compound (atomic ratio for nuclear fuel oxide Ba being 5 - 10 : 1) which is to prevent fission products breaking the protective carbon and/or silicon carbide coating; the Ba or Sr molybdate generated is to reduce the pressure of the carbon dioxide produced. Methods to manufacture such nuclear fuel particles are proposed where 1) an agglomerisation and shaping of the spheres in a fast cycling bowle and 2) a formation of drops from a colloidal solution which are made to congeal in a liquid paraffin column, take place followed by the pyrolytic coating of the particles. (UWI/LH) [de

  16. Pyrometric fuel particle measurements in pressurised reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hernberg, R; Joutsenoja, T [Tampere Univ. of Technology (Finland)

    1997-10-01

    A fibre-optic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurised reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverised coal particles at the pressurised entrained flow reactor of VTT Energy in Jyvaeskylae was developed and several series of measurements were made in order to study the effects of oxygen concentration (3-30 vol%) and pressure (0.2-1.0 MPa) on the particle temperature. The fuels used in the experiments were Westerholt, Polish and Goettelborn hvb coals, Gardanne lignite and Niederberg anthracite. The initial nominal fuel particle size varied in the experiments from 70 to 250 ,{mu}m and the gas temperature was typically 1173 K. For the anthracite also the effects of gas temperature (1073-1423K) and CO{sub 2} concentration (6-80 vol%) were studied. In Orleans a fibreoptic pyrometric device was installed to a pressurised thermogravimetric reactor of CNRS and the two-colour temperatures of fuel samples were measured. The fuel in the experiments was pulverised Goettelborn char. The reliability of optical temperature measurement in this particular application was analysed. In Essen a fibre-optic pyrometric technique that is capable to measure bed and fuel particle temperatures was applied to an atmospheric fluidised bed reactor of DMT. The effects of oxygen concentration (3-8 vol%) and bed temperature (1123-1193 K) on the fuel particle temperature were studied. The fuels in these were Westerholt coal and char and EBV-coal. Some results of these measurements are presented. The project belonged to EU`s Joule 2 extension research programme (contract JOU2-CT93-0331). (orig.)

  17. 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

  18. Operation Procedure of Inspection Equipment for TRISO-coated Fuel Particle

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, Y. K.; Cho, M. S.; Kim, Y. M.; Park, J. Y.; Kim, W. J.; Jeong, K. C.; Oh, S. C.; Lee, Y. W.

    2007-03-01

    TRISO-coated fuel particle for HTGR(high temperature gas cooled reactor) is composed of fuel kernel and coating layers. The kernel and coated particle are characterized by inspection processes for inspection items such as diameter of kernel, thickness, density and an-isotropy of coating layer. The coating thickness can be nondestructively measured by X-ray inspection equipment. The coating thickness as well as the sphericity can be also measured by optical inspection system as a ceramography method. The an-isotropy can be characterized by photometer. The density of coating layer can be measured by density column. The size and sphericity of particles can be measured by PSA(particle size analyzer). The thermo-chemical characteristics of kernel can be analyzed by TG/DTA(Thermogravimetric/Differential Thermal Analyzer). The inspection objective, equipment composition, operation principle, operation manual for each equipment was described in this operation procedure, which will be used for the characterization of inspection items described above

  19. Irradiation behaviors of coated fuel particles, (4)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Ogawa, Toru; Ikawa, Katsuichi; Iwamoto, Kazumi; Ishimoto, Kiyoshi

    1981-09-01

    Loose coated fuel particles prepared in confirmity to a preliminary design for the multi-purpose VHTR in fiscal 1972 - 1974 were irradiated by 73F - 12A capsule in JMTR. Main purpose for this irradiation experiment was to examine irradiation stability of the candidate TRISO coated fuel particles for the VHTR. Also the coated particles possessing low-density kernel (90%TD), highly anisotropic OLTI-PyC and ZrC coating layer were loaded with the candidate particles in this capsule. The coated particles were irradiated up to 1.5 x 10 21 n/cm 2 of fast neutron fluence (E > 0.18 MeV) and 3.2% FIMA of burnup. In the post irradiation examination it was observed that among three kinds of TRISO particles exposed to irradiation corresponding to the normal operating condition of the VHTR ones possessing poor characteristics of the coating layers did not show a good stability. The particles irradiated under abnormally high temperature condition (> 1800 0 C) revealed 6.7% of max. EOL failure fraction (95% confidence limit). Most of these particles were failed by the ameoba effect. Furthermore, among four kinds of the TRISO particles exposed to irradiation corresponding to the transient condition of the VHTR (--1500 0 C) the two showed a good stability, while the particles possessing highly anisotropic OLTI-PyC or poorly characteristic coating layers were not so good. (author)

  20. HTGR market assessment: interim report

    International Nuclear Information System (INIS)

    1979-09-01

    The purpose of this Assessment is to establish the utility perspective on the market potential of the HTGR. The majority of issues and conclusions in this report are applicable to both the HTGR-Gas Turbine (GT) and the HTGR-Steam Cycle (SC). This phase of the HTGR Market Assessment used the HTGR-GT as the reference design as it is the present focus of the US HTGR Program. A brief system description of the HTGR-GT is included in Appendix A. This initial report provides the proposed structure for conducting the HTGR Market Assessment plus preliminary analyses to establish the magnitude and nature of key factors that affect the HTGR market. The HTGR market factors and their relationship to the present HTGR Program are discussed. This report discusses two of these factors in depth: economics and water availability. The water availability situation in the US and its impact on the potential HTGR market are described. The approach for applying the HTGR within a framework of utility systems analyses is presented

  1. Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1980-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are being used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences

  2. Silver release from coated particle fuel

    International Nuclear Information System (INIS)

    Brown, P.E.; Nabielek, H.

    1977-03-01

    The fission product Ag-110 m released from coated particles can be the dominant source of radioactivity from the core of a high temperature reactor in the early stages of the reactor life and possibly limits the accessability of primary circuit components. It can be shown that silver is retained in oxide fuel by a diffusion process (but not in carbide or carbon-diluted fuel) and that silver is released through all types of pyrocarbon layers. The retention in TRISO particles is variable and seems to be mainly connected with operating temperature and silicon carbide quality. (orig.) [de

  3. Pyrometric fuel particle measurements in pressurised reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hernberg, R.; Joutsenoja, T. [Tampere Univ. of Technology (Finland)

    1996-12-01

    A fiberoptic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurized reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverized coal particles at the pressurized entrained flow reactor in Jyvaeskylae was developed and several series of measurements were made. In Orleans a fiberoptic pyrometric device was installed to a pressurised thermogravimetric reactor and the two-colour temperatures of fuel samples were measured. Some results of these measurements are presented. The project belongs to EU`s Joule 2 extension research programme. (author)

  4. Irradiation behaviors of coated fuel particles, (3)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Iwamoto, Kazumi; Ikawa, Katsuichi

    1980-07-01

    This report is concerning to the irradiation experiments of the coated fuel particles, which were performed by 72F-6A and 72F-7A capsules in JMTR. The coated particles referred to the preliminary design of VHTR were prepared for the experiments in 1972 and 1973. 72F-6A capsule was irradiated at G-10 hole of JMTR fuel zone for 2 reactor cycles, and 72F-7A capsule had been planned to be irradiated at the same irradiation hole before 72F-6A. However, due to slight leak of the gaseous fission products into the vacuum system controlling irradiation temperature, irradiation of 72F-7A capsule was ceased after 85 hrs since the beginning. In the post irradiation examination, inspection to surface appearance, ceramography, X-ray microradiography and acid leaching for the irradiated particle samples were made, and crushing strength of the two particle samples was measured. (author)

  5. 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.

  6. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

    2009-04-01

    UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

  7. 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

  8. Process options and projected mass flows for the HTGR refabrication scrap recovery system

    International Nuclear Information System (INIS)

    Tiegs, S.M.

    1979-03-01

    The two major uranium recovery processing options reviewed are (1) internal recovery of the scrap by the refabrication system and (2) transfer to and external recovery of the scrap by the head end of the reprocessing system. Each option was reviewed with respect to equipment requirements, preparatory processing, and material accountability. Because there may be a high cost factor on transfer of scrap fuel material to the reprocessing system for recovery, all of the scrap streams will be recycled internally within the refabrication system, with the exception of reject fuel elements, which will be transferred to the head end of the reprocessing system for uranium recovery. The refabrication facility will be fully remote; thus, simple recovery techniques were selected as the reference processes for scrap recovery. Crushing, burning, and leaching methods will be used to recover uranium from the HTGR refabrication scrap fuel forms, which include particles without silicon carbide coatings, particles with silicon carbide coatings, uncarbonized fuel rods, carbon furnace parts, perchloroethylene distillation bottoms, and analytical sample remnants. Mass flows through the reference scrap recovery system were calculated for the HTGR reference recycle facility operating with the highly enriched uranium fuel cycle. Output per day from the refabrication scrap recovery system is estimated to be 4.02 kg of 2355 U and 10.85 kg of 233 U. Maximum equipment capacities were determined, and future work will be directed toward the development and costing of the scrap recovery system chosen as reference

  9. Automatic size analysis of coated fuel particles

    International Nuclear Information System (INIS)

    Wallisch, K.; Koss, P.

    1977-01-01

    The determination of the diameter, coating thickness, and sphericity of coated fuel particles by conventional methods is very time consuming. Therefore, statistical data can only be obtained with limited accuracy. An alternative method is described that avoids these disadvantages by utilizing a fast optical data-collecting system of high accuracy. This system allows the determination of the diameter of particles in the range between 100 and 1500 μm, with an accuracy of better than +-2 μm and with a rate of 100 particles per second. The density and thickness of coating layers can be determined by comparing the data obtained before and after coating, taking into account the relative increase of weight. A special device allows the automatic determination of the sphericity of single particles as well as the distribution in a batch. This device measures 50 to 100 different diameters of each particle per second. An on-line computer stores the measured data and calculates all parameters required, e.g., number of particles measured, particle diameter, standard deviation, diameter limiting values, average particle volume, average particle surface area, and the distribution of sphericity in absolute and percent form

  10. National HTGR safety program

    International Nuclear Information System (INIS)

    Davis, D.E.; Kelley, A.P. Jr.

    1982-01-01

    This paper presents an overview of the National HTGR Program in the US with emphasis on the safety and licensing strategy being pursued. This strategy centers upon the development of an integrated approach to organizing and classifying the functions needed to produce safe and economical nuclear power production. At the highest level, four plant goals are defined - Normal Operation, Core and Plant Protection, Containment Integrity and Emergency Preparedness. The HTGR features which support the attainment of each goal are described and finally a brief summary is provided of the current status of the principal safety development program supporting the validation of the four plant goals

  11. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1980

    International Nuclear Information System (INIS)

    1980-11-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an LEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbines and process heat plants

  12. Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

    International Nuclear Information System (INIS)

    Angelini, P.; Rushton, J.E.

    1979-01-01

    The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for 235 U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods

  13. Method of producing encapsulated thermonuclear fuel particles

    International Nuclear Information System (INIS)

    Smith, W.H.; Taylor, W.L.; Turner, H.L.

    1976-01-01

    A method of producing a fuel particle is disclosed, which comprises forming hollow spheroids which have a mass number greater than 50, immersing said spheroids while under the presence of pressure and heat in a gaseous atmosphere containing an isotope, such as deuterium and tritium, so as to diffuse the gas into the spheroid and thereafter cooling said spheroids up to about 77 0 Kelvin to about 4 0 Kelvin. 4 Claims, 3 Drawing Figures

  14. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  15. Computer simulation of radiation damage in HTGR elements and structural materials

    International Nuclear Information System (INIS)

    Gann, V.V.; Gurin, V.A.; Konotop, Yu.F.; Shilyaev, B.A.; Yamnitskij, V.A.

    1980-01-01

    The problem of mathematical simulation of radiation damages in material and items of HTGR is considered. A system-program complex IMITATOR, intended for imitation of neutron damages by means of charged particle beams, is used. Account of material composite structure and certain geometry of items permits to calculate fields of primary radiation damages and introductions of reaction products in composite fuel elements, microfuel elements, their shells, composite absorbing elements on the base of boron carbide, structural steels and alloys. A good correspondence of calculation and experimental burn-out of absorbing elements is obtained, application of absorbing element as medium for imitation experiments is grounded [ru

  16. Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D., E-mail: hunnjd@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-09-15

    Highlights: • Cesium release was used to detect SiC failure in HTGR fuel. • Tristructural-isotropic particles with SiC failure were isolated by gamma screening. • SiC failure was studied by X-ray tomography and SEM. • SiC degradation was observed after irradiation and subsequent safety testing. - Abstract: As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of {sup 134}Cs and {sup 137}Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were

  17. Modeling of coated fuel particles irradiation behavior

    International Nuclear Information System (INIS)

    Liang Tongxiang; Phelip, M.

    2006-01-01

    In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10 -7 . As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

  18. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  19. Analysis of Advanced Fuel Kernel Technology

    International Nuclear Information System (INIS)

    Oh, Seung Chul; Jeong, Kyung Chai; Kim, Yeon Ku; Kim, Young Min; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung

    2010-03-01

    The reference fuel for prismatic reactor concepts is based on use of an LEU UCO TRISO fissile particle. This fuel form was selected in the early 1980s for large high-temperature gas-cooled reactor (HTGR) concepts using LEU, and the selection was reconfirmed for modular designs in the mid-1980s. Limited existing irradiation data on LEU UCO TRISO fuel indicate the need for a substantial improvement in performance with regard to in-pile gaseous fission product release. Existing accident testing data on LEU UCO TRISO fuel are extremely limited, but it is generally expected that performance would be similar to that of LEU UO 2 TRISO fuel if performance under irradiation were successfully improved. Initial HTGR fuel technology was based on carbide fuel forms. In the early 1980s, as HTGR technology was transitioning from high-enriched uranium (HEU) fuel to LEU fuel. An initial effort focused on LEU prismatic design for large HTGRs resulted in the selection of UCO kernels for the fissile particles and thorium oxide (ThO 2 ) for the fertile particles. The primary reason for selection of the UCO kernel over UO 2 was reduced CO pressure, allowing higher burnup for equivalent coating thicknesses and reduced potential for kernel migration, an important failure mechanism in earlier fuels. A subsequent assessment in the mid-1980s considering modular HTGR concepts again reached agreement on UCO for the fissile particle for a prismatic design. In the early 1990s, plant cost-reduction studies led to a decision to change the fertile material from thorium to natural uranium, primarily because of a lower long-term decay heat level for the natural uranium fissile particles. Ongoing economic optimization in combination with anticipated capabilities of the UCO particles resulted in peak fissile particle burnup projection of 26% FIMA in steam cycle and gas turbine concepts

  20. Study on commercial HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo

    2000-07-01

    The Japanese energy demand in 2030 will increase up to 117% in comparison with one in 2000. We have to avoid a large consumption of fossil fuel that induces a large CO 2 emission from viewpoint of global warming. Furthermore new energy resources expected to resolve global warming have difficulty to be introduced more because of their low energy density. As a result, nuclear power still has a possibility of large introduction to meet the increasing energy demand. On the other hand, in Japan, 40% of fossil fuels in the primary energy are utilized for power generation, and the remaining are utilized as a heat source. New clean energy is required to reduce the consumption of fossil fuels and hydrogen is expected as a alternative energy resource. Prediction of potential hydrogen demand in Japan is carried out and it is clarified that the demand will potentially increase up to 4% of total primary energy in 2050. In present, steam reforming method is the most economical among hydrogen generation processes and the cost of hydrogen production is about 7 to 8 yen/m 3 in Europe and the United States and about 13 yen/m 3 in Japan. JAERI has proposed for using the HTGR whose maximum core outlet temperature is at 950degC as a heat source in the steam reforming to reduced the consumption of fossil fuels and resulting CO 2 emission. Based on the survey of the production rate and the required thermal energy in conventional industry, it is clarified that a hydrogen production system by the steam reforming is the best process for the commercial HTGR nuclear heat utilization. The HTGR steam reforming system and other candidate nuclear heat utilization systems are considered from viewpoint of system layout and economy. From the results, the hydrogen production cost in the HTGR stream reforming system is expected to be about 13.5 yen/m 3 if the cost of nuclear heat of the HTGR is the same as one of the LWR. (author)

  1. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

  2. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

  3. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

  4. Small demonstration HTGR concept

    International Nuclear Information System (INIS)

    Kiryushin, A.I.

    1989-01-01

    Currently the USSR is investigating two high-temperature gas-cooled reactors. The first plant is the VGM, a modular type HTGR with power rating of 180-250 MWth. The second plant is the VG-400 with 1000 MWth and a prestressed concrete reactor vessel. The paper contains the description of the VGM design and its main components. (author). 1 fig., 1 tab

  5. Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-01-01

    The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)

  6. Coating Thickness Measurement of the Simulated TRISO-Coated Fuel Particles using an Image Plate and a High Resolution Scanner

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Kim, Yeon Ku; Jeong, Kyung Chai; Lee, Young Woo; Kim, Bong Goo; Eom, Sung Ho; Kim, Young Min; Yeo, Sung Hwan; Cho, Moon Sung

    2014-01-01

    In this study, the thickness of the coating layers of 196 coated particles was measured using an Image Plate detector, high resolution scanner and digital image processing techniques. The experimental results are as follows. - An X-ray image was acquired for 196 simulated TRISO-coated fuel particles with ZrO 2 kernel using an Image Plate with high resolution in a reduced amount of time. - We could observe clear boundaries between coating layers for 196 particles. - The geometric distortion error was compensated for the calculation. - The coating thickness of the TRISO-coated fuel particles can be nondestructively measured using X-ray radiography and digital image processing technology. - We can increase the number of TRISO-coated particles to be inspected by increasing the number of Image Plate detectors. A TRISO-coated fuel particle for an HTGR (high temperature gas-cooled reactor) is composed of a nuclear fuel kernel and outer coating layers. The coating layers consist of buffer PyC (pyrolytic carbon), inner PyC (I-PyC), SiC, and outer PyC (O-PyC) layer. The coating thickness is measured to evaluate the soundness of the coating layers. X-ray radiography is one of the nondestructive alternatives for measuring the coating thickness without generating a radioactive waste. Several billion particles are subject to be loaded in a reactor. A lot of sample particles should be tested as much as possible. The acquired X-ray images for the measurement of coating thickness have included a small number of particles because of the restricted resolution and size of the X-ray detector. We tried to test many particles for an X-ray exposure to reduce the measurement time. In this experiment, an X-ray image was acquired for 196 simulated TRISO-coated fuel particles using an image plate and high resolution scanner with a pixel size of 25Χ25 μm 2 . The coating thickness for the particles could be measured on the image

  7. The design of cermet fuel phase fraction and fuel particle diameter

    International Nuclear Information System (INIS)

    Tian Sheng.

    1986-01-01

    UO 2 -Zr-2 is an ideal cermet fuel. As an exemplification with this fuel, this paper emphatically elucidates the irradiation theory of cermet fuel and its application in the design of cermet fuel phase fraction and of fuel particle diameter. From the point of view of the irradiation theory and the consideration for sandwich rolling, the suitable volume fraction of UO 2 phase of 25% and diameter of UO 2 particle of 100 +- 15 μm are selected

  8. Particle fueling and impurity control in PDX

    International Nuclear Information System (INIS)

    Fonck, R.J.; Bell, M.; Bol, K.

    1984-12-01

    Fueling requirements and impurity levels in neutral-beam-heated discharges in the PDX tokamak have been compared for plasmas formed with conventional graphite rail limiters, a particle scoop limiter, and an open or closed poloidal divertor. Gas flows necessary to obtain a given density are highest for diverted discharges and lowest for the scoop limiter. Hydrogen pellet injection provides an efficient alternate fueling technique, and a multiple pellet injector has produced high density discharges for an absorbed neutral beam power of up to 600 kW, above which higher speeds or more massive pellets are required for penetration to the plasma core. Power balance studies indicate that 30 to 40% of the total input power is radiated while approx. 15% is absorbed by the limiting surface, except in the open divertor case, where 60% flows to the neutralizer plate. In all operating configurations, Z/sub eff/ usually rises at the onset of neutral beam injection. Both open divertor plasmas and those formed on a well conditioned water-cooled limiter have Z/sub eff/ less than or equal to 2 at the end of neutral injection. A definitive comparison of divertors and limiters for impurity control purposes requires longer beam pulses or higher power levels than available on present machines

  9. HTGR depressurization analysis

    International Nuclear Information System (INIS)

    Boccio, J.L.; Colman, J.; Skalyo, J.; Beerman, J.

    1979-01-01

    Relaxation of the prima facie assumption of complete mixing of primary and secondary containment gases during HTGR depressurization has led to a study program designed to identify and selectively quantify the relevant gas dynamic processes which prevail during the depressurization event. Uncertainty in the degree of gas mixedness naturally leads to uncertainty in containment vessel design pressure and heat loads and possible combustion hazards therein. This paper succinctly details an analytical approach and modeling methodology of the exhaust jet structure/containment vessel interaction during penetration failures. (author)

  10. Processing of FRG mixed oxide 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.; Tischer, H.E.

    1980-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing 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. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures

  11. Steam generator design considerations for modular HTGR plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; DeFur, D.D.

    1986-01-01

    Studies are in progress to develop a standard High Temperature Gas-Cooled Reactor (HTGR) plant design that is amenable to serial production and is licensable. Based on the results of trade studies performed in the DOE-funded HTGR program, activities are being focused to emphasize a modular concept based on a 350 MW(t) annular reactor core with prismatic fuel elements. Utilization of a multiplicity of the standard module affords flexibility in power rating for utility electricity generation. The selected modular HTGR concept has the reactor core and heat transport systems housed in separate steel vessels. This paper highlights the steam generator design considerations for the reference plant, and includes a discussion of the major features of the heat exchanger concept and the technology base existing in the U.S

  12. Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

    International Nuclear Information System (INIS)

    Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

    1977-02-01

    Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

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

    Energy Technology Data Exchange (ETDEWEB)

    Bosser, R; Langlet, G

    1972-06-15

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

  14. Management feature of transuranic for HTGR and LWR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Long-lived actinides from spent fuels can cause potential long-term environ- mental hazards. The generation and incineration of transuranic in different closed fuel cycles were studied. U and Pu were recycled from spent fuel in the 250 MW high-temperature gas-cooled reactor-pebble-bed-module (HTR-PM) U-Pu fuelled core, and then PuO 2 and MOX fuel elements were designed based on this recycled U and Pu. These fuel elements were used to build up a new PuO 2 or MOX fuelled core with the same geometry of the original reactor. Characteristics of transuranic incineration with HTGR open and closed fuel cycles were studied with VSOP code, and the corresponding results from the light water reactor were compared and analyzed. The transuranic generation with HTGR open fuel cycle is almost half of the corresponding result of the light water reactor. Thus, HTGR closed fuel cycles can effectively burn transuranic. (authors)

  15. HTGR Economic / Business Analysis and Trade Studies Market Analysis for HTGR Technologies and Applications

    Energy Technology Data Exchange (ETDEWEB)

    Richards, Matt [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States); Hamilton, Chris [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States)

    2013-11-01

    This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal to liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.

  16. Comparative study of random and uniform models for the distribution of TRISO particles in HTR-10 fuel elements

    International Nuclear Information System (INIS)

    Rosales, J.; Perez, J.; Garcia, C.; Munnoz, A.; Lira, C. A. B. O.

    2015-01-01

    TRISO particles are the specific features of HTR-10 and generally HTGR reactors. Their heterogeneity and random arrangement in graphite matrix of these reactors create a significant modeling challenge. In the simulation of spherical fuel elements using MCNPX are usually created repetitive structures using uniform distribution models. The use of these repetitive structures introduces two major approaches: the non-randomness of the TRISO particles inside the pebbles and the intersection of the pebble surface with the TRISO particles. These approaches could affect significantly the multiplicative properties of the core. In order to study the influence of these approaches in the multiplicative properties was estimated the K inf value in one pebble with white boundary conditions using 4 different configurations regarding the distribution of the TRISO particles inside the pebble: uniform hexagonal model, cubic uniform model, cubic uniform without the effect of cutting and a random distribution model. It was studied the impact these models on core scale solving the problem B1, from the Benchmark Problems presented in a Coordinated Research Program of the IAEA. (Author)

  17. Status of international HTGR development

    International Nuclear Information System (INIS)

    Homan, F.J.; Simon, W.A.

    1988-01-01

    Programs for the development of high-temperature gas-cooled reactor (HTGR) technology over the past 30 years in eight countries are briefly described. These programs have included both government sector and industrial sector participation. The programs have produced four electricity-producing prototype/demonstration reactors, two in the United States, and two in the Federal Republic of Germany. Key design parameters for these ractors are compared with the design parameters planned for follow-on commercial-scale HTGRs. The development of HTGR technology has been enhanced by numerous cooperative agreements over the years, involving both government-sponsored national laboratories and industrial participants. Current bilateral cooperative agreements are described. A relatively new component in the HTGR international cooperation is that of multinational industrial alliances focused on supplying commercial-scale HTGR power plants. Current industrial cooperative agreements are briefly discussed

  18. HTGR Cost Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooler Reactor (HTGR) Cost Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Cost Model calculates an estimate of the capital costs, annual operating and maintenance costs, and decommissioning costs for a high-temperature gas-cooled reactor. The user can generate these costs for multiple reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for a single or four-pack configuration; and for a reactor size of 350 or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Cost Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Cost Model. This model was design for users who are familiar with the HTGR design and Excel. Modification of the HTGR Cost Model should only be performed by users familiar with Excel and Visual Basic.

  19. Effects of the HTGR-gas turbine on national reactor strategies

    International Nuclear Information System (INIS)

    Ligon, D.M.; Brogli, R.H.

    1979-11-01

    A specific role for the HTGR in a national energy strategy is examined. The issue is addressed in two ways. First, the role of the HTGR-GT Binary cycle plant is examined in a national energy strategy based on symbiosis between fast breeder and advanced converter reactors utilizing the thorium U233 fuel cycle. Second, the advantages of the HTGR-GT dry-cooled plant operating in arid regions is examined and compared with a dry-cooled LWR. An event tree analysis of potential benefits is applied

  20. Utilization of particle fuels in different reactor concepts

    International Nuclear Information System (INIS)

    1983-04-01

    To date, particle fuel is only used in high temperature reactors (HTR). In this reactor type the particles exist of oxide fuel with a diameter of about 0.5 mm and are surrounded by various coatings in order to safely enclose fission products and decrease the radioactive release into the primary circuit. However, it is felt that fuel based upon spherical particles could have some advantages compared with pellets both on fabrication and in-core behaviour in several reactor concepts. This fuel is now of general interest and there is a high level of research and development activity in some countries. In order to collect, organize additional information and summarize experience on utilization of particle fuels in different reactor concepts, a questionnaire was prepared by IAEA in 1980 and sent to Member States, which might be involved in relevant developments. This survey has been prepared by a group of consultants and is mainly based on the responses to the IAEA questionnaire

  1. Critical Issues for Particle-Bed Reactor Fuels

    Science.gov (United States)

    Evans, Robert S.; Husser, Dewayne L.; Jensen, Russell R.; Kerr, John M.

    1994-07-01

    Particle-Bed Reactors (PBRs) potentially offer performance advantages for nuclear thermal propulsion, including very high power densities, thrust-to-weight ratios, and specific impulses. A key factor in achieving all of these is the development of a very-high-temperature fuel. The critical issues for all such PBR fuels are uranium loading, thermomechanical and thermochemical stability, compatibility with contacting materials, fission product retention, manufacturability, and operational tolerance for particle failures. Each issue is discussed with respect to its importance to PBR operation, its status among current fuels, and additional development needs. Mixed-carbide-based fuels are recommended for further development to support high-performance PBRs.

  2. Evaluation of Particle Counter Technology for Detection of Fuel Contamination Detection Utilizing Fuel System Supply Point

    Science.gov (United States)

    2014-06-19

    product used as a diesel product for ground use (1). Free water contamination (droplets) may appear as fine droplets or slugs of water in the fuel...methods and test procedures for the calibration and use of automatic particle counters. The transition of this technology to the fuel industry is...UNCLASSIFIED 6 UNCLASSIFIED Receipt Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18

  3. HTGR safety research concerns at NRC

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1982-01-01

    A general discussion of HTGR technical and safety-related problems is given. The broad areas of current research programs specific to the Fort St. Vrain reactor and applicable to HTGR technology are summarized

  4. Status of CHAP: composite HTGR analysis program

    International Nuclear Information System (INIS)

    Secker, P.A.; Gilbert, J.S.

    1975-12-01

    Development of an HTGR accident simulation program is in progress for the prediction of the overall HTGR plant transient response to various initiating events. The status of the digital computer program named CHAP (Composite HTGR Analysis Program) as of June 30, 1975, is given. The philosophy, structure, and capabilities of the CHAP code are discussed. Mathematical descriptions are given for those HTGR components that have been modeled. Component model validation and evaluation using auxiliary analysis codes are also discussed

  5. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  6. USNRC HTGR safety research program overview

    International Nuclear Information System (INIS)

    Foulds, R.B.

    1982-01-01

    An overview is given of current activities and planned research efforts of the US Nuclear Regulatory Commission (NRC) HTGR Safety Program. On-going research at Brookhaven National Laboratory, Oak Ridge National Laboratory, Los Alamos National Laboratory, and Pacific Northwest Laboratory are outlined. Tables include: HTGR Safety Issues, Program Tasks, HTGR Computer Code Library, and Milestones for Long Range Research Plan

  7. HTGR analytical methods and design verification

    International Nuclear Information System (INIS)

    Neylan, A.J.; Northup, T.E.

    1982-05-01

    Analytical methods for the high-temperature gas-cooled reactor (HTGR) include development, update, verification, documentation, and maintenance of all computer codes for HTGR design and analysis. This paper presents selected nuclear, structural mechanics, seismic, and systems analytical methods related to the HTGR core. This paper also reviews design verification tests in the reactor core, reactor internals, steam generator, and thermal barrier

  8. Prospects of HTGR process heat application and role of HTTR

    International Nuclear Information System (INIS)

    Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    At Japan Atomic Energy Research Institute, an effort on development of process heat application with high temperature gas cooled reactor (HTGR) has been continued for providing a future clean alternative to the burning of fossil energy for the production of industrial process heat. The project is named 'HTTR Heat Utilization Project', which includes a demonstration of hydrogen production using the first Japanese HTGR of High Temperature Engineering Test Reactor (HTTR). In the meantime, some countries, such as China, Indonesia, Russia and South Africa are trying to explore the HTGR process heat application for industrial use. One of the key issues for this application is economy. It has been recognized for a long time and still now that the HTGR heat application system is not economically competitive to the current fossil ones, because of the high cost of the HTGR itself. However, the recent movement on the HTGR development, as represented by South Africa Pebble Beds Modular Reactor (SA-PBMR) Project, has revealed that the HTGRs are well economically competitive in electricity production to fossil fuel energy supply under a certain condition. This suggests that the HTGR process heat application will also possibly get economical in the near future. In the present paper, following a brief introduction describing the necessity of the HTGRs for the future process heat application, Japanese activities and prospect of the development on the process heat application with the HTGRs are described in relation with the HTTR Project. In conclusion, the process heat application system with HTGRs is thought technically and economically to be one of the most promising applications to solve the global environmental issues and energy shortage which may happen in the future. However, the commercialization for the hydrogen production system from water, which is the final goal of the HTGR process heat application, must await the technology development to be completed in 2030's at the

  9. Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.

    1975-01-01

    In the FRG three basic fuel element designs for application in high temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margin and presents aspects of their future development potential for operation in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. Summarizing it can be said that all the fuel elements as well as their components have proved their aptitude for the dual cycle systems in numerous fuel element and particle performance tests. To adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity further testing is still necessary, however. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.) [de

  10. Physicochemical analysis of interaction of oxide fuel with pyrocarbon coatings of fuel particles

    International Nuclear Information System (INIS)

    Lyutikov, R.A.; Khromov, Yu.F.; Chernikov, A.S.

    1990-01-01

    Equilibrium pressure of (CO+Kr,Xe) gases inside fuel particle with oxide kern depending on design features of fuel particle, on temperature. on (O/U) initial composition and fuel burnup is calculated using the suggested model. Analysis of possibility for gas pressure reduction by means of uranium carbide alloying of kern and degree increase of solid fission product retention (Cs for example) during alumosilicate alloying of uranium oxide is conducted

  11. Effect of deposition conditions on the properties of pyrolytic silicon carbide coatings for high-temperature gas-cooled reactor fuel particles

    International Nuclear Information System (INIS)

    Stinton, D.P.; Lackey, W.J.

    1977-10-01

    Silicon carbide coatings on HTGR microsphere fuel act as the barrier to contain metallic fission products. Silicon carbide coatings were applied by the decomposition of CH 3 SiCl 3 in a 13-cm-diam (5-in.) fluidized-bed coating furnace. The effects of temperature, CH 3 SiCl 3 supply rate and the H 2 :CH 3 SiCl 3 ratio on coating properties were studied. Deposition temperature was found to control coating density, whole particle crushing strength, coating efficiency, and microstructure. Coating density and microstructure were also partially determined by the H 2 :CH 3 SiCl 3 ratio. From this work, it appears that the rate at which high quality SiC can be deposited can be increased from 0.2 to 0.5 μm/min

  12. Personnel radiation exposure in HTGR plants

    International Nuclear Information System (INIS)

    Su, S.; Engholm, B.A.

    1981-01-01

    Occupational radiation exposures in high-temperature gas-cooled reactor (HTGR) plants were assessed. The expected rate of dose accumulations for a large HTGR steam cycle unit is 0.07 man-rem/MW(e)y, while the design basis is 0.17 man-rem/MW(e)y. The comparable figure for actual light water reactor experience is 1.3 man-rem/MW(e)y. The favorable HTGR occupational exposure is supported by results from the Peach Bottom Unit No. 1 HTGR and Fort St. Vrain HTGR plants and by operating experience at British gas-cooled reactor stations

  13. Developmental assessment of the Fort St. Vrain version of the composite HTGR analysis program (CHAP-2)

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1981-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are being used to partially verify the component modeling and dynamic simulation techniques used to predict plant response to postulated accident sequences. Results of these preliminary validation efforts are presented showing good agreement between code output and plant data for the portions of the code that have been tested. Plans for further development and assessment as well as application of the validated code are discussed. (author)

  14. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  15. Improved graphite matrix for coated-particle fuel

    International Nuclear Information System (INIS)

    Schell, D.H.; Davidson, K.V.

    1978-10-01

    An experimental process was developed to incorporate coated fuel particles in an extruded graphite matrix. This structure, containing 41 vol% particles, had a high matrix density, >1.6 g/cm 3 , and a matrix conductivity three to four times that of a pitch-injected fuel rod at 1775 K. Experiments were conducted to determine the uniformity of particle loadings in extrusions. Irradiation specimens were supplied for five tests in the High-Fluence Isotope Reactor at the Oak Ridge National Laboratory

  16. Fueling profile sensitivities of trapped particle mode transport to TNS

    International Nuclear Information System (INIS)

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

    1977-01-01

    A key factor in the plasma thermal behavior is the anticipated existence of dissipative trapped particle modes. A possible scheme for controlling the strength of these modes was found. The scheme involves varying the cold fueling profile. A one dimensional multifluid transport code was used to simulate plasma behavior. A multiregime model for particle and energy transport was incorporated based on pseudoclassical, trapped electron, and trapped ion regimes used elsewhere in simulation of large tokamaks. Fueling profiles peaked toward the plasma edge may provide a means for reducing density-gradient-driven trapped particle modes, thus reducing diffusion and conduction losses

  17. Spot Ignition of Natural Fuels by Hot Metal Particles

    OpenAIRE

    Urban, James Linwood

    2017-01-01

    The spot ignition of combustible material by hot metal particles is an important pathway by which wildland and urban spot fires and smolders are started. Upon impact with a fuel, such as dry grass, duff, or saw dust, these particles can initiate spot fires by direct flaming or smoldering which can transition to a flame. These particles can be produced by processes such as welding, powerline interactions, fragments from bullet impacts, abrasive cutting, and pyrotechnics. There is little publi...

  18. Summary of foreign HTGR programs

    International Nuclear Information System (INIS)

    1980-06-01

    This report contains pertinent information on the status, objectives, budgets, major projects and facilities, as well as user, industrial and governmental organizations involved in major foreign gas-cooled thermal reactor programs. This is the second issue of this document (the first was issued in March 1979). The format has been revised to consolidate material according to country. These sections are followed by the foreign HTGR program index which serves as a quick reference to some of the many acronyms associated with the foreign HTGR programs

  19. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2008

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Tachibana, Yukio; Sun Yuliang

    2009-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2008. (author)

  20. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2009

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Wang Hong

    2010-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2009. (author)

  1. Information exchange mainly on HTGR operation and maintenance technique between JAEA and INET in 2005

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Hino, Ryutaro; Yu Suyuan

    2006-06-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation with emphasis on HTGR operation and maintenance techniques between JAEA and INET and outlines cooperation activities during the fiscal year 2005. (author)

  2. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

    International Nuclear Information System (INIS)

    Ding Shurong; Wang Qiming; Huo Yongzhong

    2010-01-01

    In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

  3. Is there a chance for commercializing the HTGR in Indonesia?

    International Nuclear Information System (INIS)

    Arbie, B.; Akhmad, Y.R.

    1997-01-01

    Indonesia is one of the developing countries in Asia-Pacific regions that actively improving or at least continuously maintain its economic growth. For this purpose, to fulfill a domestic energy demand is a vital role to achieve the goals of Indonesian development. Pertamina, the state-owned oil company, has recently called for a significant increase in domestic gas consumption in a bid to delay Indonesia becoming a net oil importer. Therefore, there is good chance for gas industry to increase their roles in generating electricity and producing automotive fuels. The latter is an interesting field of study to be correlated with the utilization of HTGR technology where the heat source could be used in the reforming process to convert natural gas into syngas as feed material in producing automotive fuels. Since the end of 1995 National Atomic Energy Agency of Indonesia (BATAN) has made an effort to increase its role in the national energy program and Batan is also able to revolve in the Giant Natuna Project or the other natural gas field projects to promote syngas production applying HTGR technology. A series of meeting with Pertamina and BPPT (the Agency for the Assessment and Application of Technology) had been performed to promote utilization of HTGR technology in the Natuna Project. In this paper governmental policy for natural gas production that may closely relate to syngas production and preliminary study for production of syngas at the Natuna Project will be discussed. It is concluded that to gain the possibility of the HTGR acceptance in the project a scenario for production and distribution should be arranged in other to achieve the break even point for automotive fuel price at about 10 US$/GJ (fuel price in 1996) in Indonesia. (author)

  4. Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

    International Nuclear Information System (INIS)

    Long, Jr. E.L.

    2001-01-01

    Seven full-sized Peach Bottom Reactor fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10 21 neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10 21 , but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum

  5. HTGR accident and risk assessment

    International Nuclear Information System (INIS)

    Silady, F.A.; Everline, C.J.; Houghton, W.J.

    1982-01-01

    This paper is a synopsis of the high-temperature gas-cooled reactor probabilistic risk assessments (PRAs) performed by General Atomic Company. Principal topics presented include: HTGR safety assessments, peer interfaces, safety research, process gas explosions, quantitative safety goals, licensing applications of PRA, enhanced safety, investment risk assessments, and PRA design integration

  6. Present status of research on hydrogen energy and perspective of HTGR hydrogen production system

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Ogawa, Masuro; Akino, Norio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-03-01

    A study was performed to make a clear positioning of research and development on hydrogen production systems with a High Temperature Gas-cooled Reactor (HTGR) under currently promoting at the Japan Atomic Energy Research Institute through a grasp of the present status of hydrogen energy, focussing on its production and utilization as an energy in future. The study made clear that introduction of safe distance concept for hydrogen fire and explosion was practicable for a HTGR hydrogen production system, including hydrogen properties and need to provide regulations applying to handle hydrogen. And also generalization of hydrogen production processes showed technical issues of the HTGR system. Hydrogen with HTGR was competitive to one with fossil fired system due to evaluation of production cost. Hydrogen is expected to be used as promising fuel of fuel cell cars in future. In addition, the study indicated that there were a large amount of energy demand alternative to high efficiency power generation and fossil fuel with nuclear energy through the structure of energy demand and supply in Japan. Assuming that hydrogen with HTGR meets all demand of fuel cell cars, an estimation would show introduction of the maximum number of about 30 HTGRs with capacity of 100 MWt from 2020 to 2030. (author)

  7. HTGR strategy for reduced proliferation potential

    International Nuclear Information System (INIS)

    Stewart, H.B.; Dahlberg, R.C.

    1978-01-01

    The HTGR stratregy for reduced proliferation potential is one aspect of a potential broader nuclear strategy aimed primarily toward a transition nuclear period between today's uranium-consumption reactors and the long-range balanced system of breeder and advanced near-breeder reactors. In particular, the normal commerce of U-233 could be made acceptable by: (a) dependence on the gamma radiation from U-232 daughter products, (b) enhancement of that radioactivity by incomplete fission-product decontamination of the bred-fuel, or (c) denaturing of the U-233 with U-238. These approaches would, of course, supplement institutional initiatives to improve proliferation resistance such as the collocation of facilities and the establishment of secure energy centers. 6 refs

  8. HTGR-GT systems optimization studies

    International Nuclear Information System (INIS)

    Kammerzell, L.L.; Read, J.W.

    1980-06-01

    The compatibility of the inherent features of the high-temperature gas-cooled reactor (HTGR) and the closed-cycle gas turbine combined into a power conversion system results in a plant with characteristics consistent with projected utility needs and national energy goals. These characteristics are: (1) plant siting flexibility; (2) high resource utilization; (3) low safety risks; (4) proliferation resistance; and (5) low occupational exposure for operating and maintenance personnel. System design and evaluation studies on dry-cooled intercooled and nonintercooled commercial plants in the 800-MW(e) to 1200-MW(e) size range are described, with emphasis on the sensitivity of plant design objectives to variation of component and plant design parameters. The impact of these parameters on fuel cycle, fission product release, total plant economics, sensitivity to escalation rates, and plant capacity factors is examined

  9. Fission product released experiment of coated fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

    1992-01-15

    Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

  10. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

    2016-10-15

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  11. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    International Nuclear Information System (INIS)

    Tavron, Barak; Shwageraus, Eugene

    2016-01-01

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  12. Peach Bottom HTGR decommissioning and component removal

    International Nuclear Information System (INIS)

    Kohler, E.J.; Steward, K.P.; Iacono, J.V.

    1977-07-01

    The prime objective of the Peach Bottom End-of-Life Program was to validate specific HTGR design codes and predictions by comparison of actual and predicted physics, thermal, fission product, and materials behavior in Peach Bottom. Three consecutive phases of the program provide input to the HTGR design methods verifications: (1) Nondestructive fuel and circuit gamma scanning; (2) removal of steam generator and primary circuit components; and (3) Laboratory examinations of removed components. Component removal site work commenced with establishment of restricted access areas and installation of controlled atmosphere tents to retain relative humidity at <30%. A mock-up room was established to test and develop the tooling and to train operators under simulated working conditions. Primary circuit ducting samples were removed by trepanning, and steam generator access was achieved by a combination of arc gouging and grinding. Tubing samples were removed using internal cutters and external grinding. Throughout the component removal phase, strict health physics, safety, and quality assurance programs were implemented. A total of 148 samples of primary circuit ducting and steam generator tubing were removed with no significant health physics or safety incidents. Additionally, component removal served to provide access fordetermination of cesium plateout distribution by gamma scanning inside the ducts and for macroexamination of the steam generator from both the water and helium sides. Evaluations are continuing and indicate excellent performance of the steam generator and other materials, together with close correlation of observed and predicted fission product plateout distributions. It is concluded that such a program of end-of-life research, when appropriately coordinated with decommissioning activities, can significantly advance nuclear plant and fuel technology development

  13. The commercial application prospect of HTGR plant in China

    International Nuclear Information System (INIS)

    Wang Yingsu

    2008-01-01

    With an introduction of the features and current situation of the HTGR power generation as well as the development of HTGR demonstration project in China, the article analyzes the necessity of developing HTGR power plants. The article proposes to exercise the advantages of HTGR to full extent so as to further develop HTGR power plants. It is believed that HTGR is of great commercial promotion value under appropriate circumstances. (authors)

  14. Nuclear heat source design for an advanced HTGR process heat plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; O'Hanlon, T.W.

    1983-01-01

    A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

  15. Light Obscuration Particle Counter Fuel Contamination Limits

    Science.gov (United States)

    2015-10-08

    two methods for measuring particulate contamination by gravimetric analysis: ASTM D2276 - Standard Test Method for Particulate Contaminant in Aviation...problems with the gravimetric methods is the poor repeatability and reproducibility of the methods , ASTM D2276 has a repeatability of 0.25 mg/L and...collected failing either the particle count or the gravimetric limit, 314 of which failed both methods . 245 false negatives were recorded where the

  16. Coated fuel particles: requirements and status of fabrication technology

    International Nuclear Information System (INIS)

    Huschka, H.; Vygen, P.

    1977-01-01

    Fuel cycle, design, and irradiation performance requirements impose restraints on the fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high-temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. Quality reliability of coated fuel particles is ensured by quality control and quality assurance programs operated by an aduiting system that includes licensing officials and the customer

  17. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

  18. 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

  19. Effects of ashes in solid fuels on fuel particle charging during combustion in an air stream

    Energy Technology Data Exchange (ETDEWEB)

    Zakharov, A.G.; Fialkov, B.S.; Mel' nichuk, A.Yu.; Khvan, L.A.

    1982-09-01

    Black coal from the Karaganda basin is mixed with sodium chloride and graphite. Coal characteristics are given in a table (density, ashes, content of silica, aluminium oxides, iron oxides, calcium oxides, potassium oxides and magnesium oxides). Effects of ash fluctuations on electric potential of fuel particles during combustion are analyzed. Analyses show that with increasing ash content electric potential of fuel particles decreases and reaches the minimum when ash content ranges from 70 to 80 %. Particles with electric potential are generated during chemical processes between carbon and oxygen when coal is burned in an air stream. (5 refs.) (In Russian)

  20. 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

  1. Graphite oxidation in HTGR atmosphere

    International Nuclear Information System (INIS)

    Growcock, F.B.; Barry, J.J.; Finfrock, C.C.; Rivera, E.; Heiser, J.H. III

    1982-01-01

    On-going and recently completed studies of the effect of thermal oxidation on the structural integrity of HTGR candidate graphites are described, and some results are presented and discussed. This work includes the study of graphite properties which may play decisive roles in the graphites' resistance to oxidation and fracture: pore size distribution, specific surface area and impurity distribution. Studies of strength loss mechanisms in addition to normal oxidation are described. Emphasis is placed on investigations of the gas permeability of HTGR graphites and the surface burnoff phenomenon observed during recent density profile measurements. The recently completed studies of catalytic pitting and the effects of prestress and stress on reactivity and ultimate strength are also discussed

  2. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  3. Improved moulding material for addition to nuclear fuel particles to produce nuclear fuel elements

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1976-01-01

    A suggestion is made to improve the moulding materials used to produce carbon-contained nuclear fuel particles by a coke-reducing added substance. The nuclear fuel particles are meant for the formation of fuel elements for gas-cooled high-temperature nuclear reactors. The moulding materials are above all for the formation of coated particles which are burnt in situ in nuclear fuel element chambers out of 'green' nuclear fuel bodies. The added substance improves the shape stability of the particles forming and prevents a stiding or bridge formation between the particles or with the surrounding walls. The following are named as added substances: 1) Polystyrene and styrene-butadiene-Co polymers (mol. wt. between 5oo and 1,000,000), 2) aromatic compounds (mol. wt. 75 to 300), 3) saturated hydrocarbon polymers (mol. wt. 5,000 to 1,000,000). Additional release agents further improve the properties in the same direction (e.g. alcohols, fatty acids, amines). (orig.) [de

  4. US HTGR Deployment Challenges and Strategies HTR 2014 Conference Proceedings

    International Nuclear Information System (INIS)

    Shahrokhi, Farshid; Lommers, Lewis; Mayer, John III; Southworth, Finis

    2014-01-01

    The NGNP Industry Alliance (NIA), LLC (www.NGNPAliance.org), is a consortium of high temperature gas-cooled reactor (HTGR) designers, utility plant owner/operators, critical plant hardware suppliers, and end-user groups. The NIA is promoting the design and commercialization of a HTGR for industrial process heat applications and electricity generation. In 2012, NIA selected the AREVA Steam Cycle HTGR (SC-HTGR) as its primary reactor design choice for its first implementation in mid -2020s. The SC-HTGR can produce 625 MWth of process steam at 550°C or 275 MWe of electricity in a co-generation configuration. The standard plant is a four-pack of 625MWth modules providing steam and electricity co-generation. The safety characteristics of the HTGR technology allows close colocation of the nuclear plant and the industrial end-user. The plant design also allows the process steam used for the industrial applications to be completely segregated and separate from primary Helium coolant and the secondary nuclear steam supply systems. The process steam at temperatures up to 550°C is provided for a variety of direct or indirect applications. End-user requirements are met for a wide range of steam flow, pressure and temperature conditions. Very high reliability (>99.99%) is maintained by the use of multi-reactor modules and conventional gas-fired back-up. Intermittent steam loads can also be efficiently met through co-generation of electricity for internal use or external distribution and sale. The NIA technology development and deployment challenges are met with strategies that provide investment and partnerships opportunities for plant design and equipment supply, and by cooperative government research, sovereign or private investment, and philanthropic opportunities. Our goal is to create intellectual property (IP) and investor value as the design matures and a license is obtained. The strategy also includes involvement of the initial customer in sharing the value created in

  5. Growth of the interaction layer around fuel particles in dispersion fuel

    International Nuclear Information System (INIS)

    Olander, D.

    2009-01-01

    Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x . The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel

  6. Uncertainties in HTGR neutron-physical characteristics due to computational errors and technological tolerances

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Grebennik, V.N.; Davidenko, V.G.; Kosovskij, V.G.; Smirnov, O.N.; Tsibul'skij, V.F.

    1991-01-01

    The paper is dedicated to the consideration of uncertainties is neutron-physical characteristics (NPC) of high-temperature gas-cooled reactors (HTGR) with a core as spherical fuel element bed, which are caused by calculations from HTGR parameters mean values affecting NPC. Among NPC are: effective multiplication factor, burnup depth, reactivity effect, control element worth, distribution of neutrons and heat release over a reactor core, etc. The short description of calculated methods and codes used for HTGR calculations in the USSR is given and evaluations of NPC uncertainties of the methodical character are presented. Besides, the analysis of the effect technological deviations in parameters of reactor main elements such as uranium amount in the spherical fuel element, number of neutron-absorbing impurities in the reactor core and reflector, etc, upon the NPC is carried out. Results of some experimental studies of NPC of critical assemblies with graphite moderator are given as applied to HTGR. The comparison of calculations results and experiments on critical assemblies has made it possible to evaluate uncertainties of calculated description of HTGR NPC. (author). 8 refs, 8 figs, 6 tabs

  7. Washing of gel particles in wet chemical manufacture of reactor fuel particles

    International Nuclear Information System (INIS)

    Ringel, H.

    1980-07-01

    In the manufacture of HTR fuel particles and particles of fertile material by wet chemical methods, the ammonium nitrate formed during the precipitation reaction must be washed out of the gel particles. This washing process has been investigated theoretically and experimentally. A counter-current washer has been developed which in particular takes account of the aspects of refabrication - such as compact construction and minimum waste. A counter-current washing column of 17 mm internal diameter and 640 mm length gives to gel particle throughput of 0.65 1/h. The volume ratio of wash water to gel particles is 5, and the residual nitrate concentration in the particles is 7 x 10 -3 mols of NO - 3 /1. (orig.) [de

  8. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  9. Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.; Detavernier, C.

    2010-01-01

    Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

  10. Corrosion Studies of Platinum Nano-Particles for Fuel Cells

    DEFF Research Database (Denmark)

    Shim, Signe Sarah

    The main focus of the present thesis is on corrosion and prevention of corrosion of platinum particles supported on carbon. This is important for instance in connection with start up and shutdown of fuel cells. The degradation mechanism of platinum particles supported on carbon has been character......The main focus of the present thesis is on corrosion and prevention of corrosion of platinum particles supported on carbon. This is important for instance in connection with start up and shutdown of fuel cells. The degradation mechanism of platinum particles supported on carbon has been...... characterized during oxygen reduction reaction (ORR) condition using identical location (IL) transmission electron microscopy (TEM). A TEM grid was used as the working electrode in an electrochemical setup allowing a direct correlation between the electrochemical response and the TEM analysis. The main results...... thirds and one monolayer of gold on platinum supported on carbon were synthesized by an inverse micelle method. The results obtained appear independent of the gold coverage. It has been shown that the electrochemical active surface areas of the platinum and platinum gold particles synthesized...

  11. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  12. Particle fueling experiments with a series of pellets in LHD

    Science.gov (United States)

    Baldzuhn, J.; Damm, H.; Dinklage, A.; Sakamoto, R.; Motojima, G.; Yasuhara, R.; Ida, K.; Yamada, H.; LHD Experiment Group; Wendelstein 7-X Team

    2018-03-01

    Ice pellet injection is performed in the heliotron Large Helical Device (LHD). The pellets are injected in short series, with up to eight individual pellets. Parameter variations are performed for the pellet ice isotopes, the LHD magnetic configurations, the heating scenario, and some others. These experiments are performed in order to find out whether deeper fueling can be achieved with a series of pellets compared to single pellets. An increase of the fueling efficiency is expected since pre-cooling of the plasma by the first pellets within a series could aid deeper penetration of later pellets in the same series. In addition, these experiments show which boundary conditions must be fulfilled to optimize the technique. The high-field side injection of pellets, as proposed for deep fueling in a tokamak, will not be feasible with the same efficiency in a stellarator or heliotron because there the magnetic field gradient is smaller than in a tokamak of comparable size. Hence, too shallow pellet fueling, in particular in a large device or a fusion reactor, will be an issue that can be overcome only by extremely high pellet velocities, or other techniques that will have to be developed in the future. It turned out by our investigations that the fueling efficiency can be enhanced by the injection of a series of pellets to some extent. However, further investigations will be needed in order to optimize this approach for deep particle fueling.

  13. Method to produce carbon-cladded nuclear fuel particles

    International Nuclear Information System (INIS)

    Sturge, D.W.; Meaden, G.W.

    1978-01-01

    In the method charges of micro-spherules of fuel element are designed to have two carbon layers, whereby a one aims to achieve a uniform granulation (standard measurement). Two drums are used for this purpose connected behind one another. The micro-spherules coated with the first layer (phenolformaldehyde resin coated graphite particles) leave the first drum and enter the second one. Following the coating with a second layer, the micro-spherules are introduced into a grain size separator. The spherules that are too small are directly recycled into the second drum and those ones that are too large are recycled into the first drum after removing the graphite layers. The method may also be applied to metal cladded particles to manufacture cermet fuels. (RW) [de

  14. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Ploger, Scott A., E-mail: scott.ploger@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Demkowicz, Paul A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Hunn, John D.; Kehn, Jay S. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2014-05-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak compact-average burnup of 19.5% FIMA with no in-pile failures observed out of 3 × 10{sup 5} total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Six compacts have been examined, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose from 36 to 79 individual particles near midplane on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer–IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, 981 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in 23% of the particles, and these fractures often resulted in unconstrained kernel protrusion into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer–IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only four classified particles, all in conjunction with IPyC–SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures and IPyC–SiC debonds.

  15. 1170-MW(t) HTGR-PS/C plant application study report: Geismar, Louisiana refinery/chemical complex application

    International Nuclear Information System (INIS)

    McMain, A.T. Jr.; Stanley, J.D.

    1981-05-01

    This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to an industrial complex at Geismar, Louisiana. This study compares the HTGR with coal and oil as process plant fuels. This study uses a previous broad energy alternative study by the Stone and Webster Corporation on refinery and chemical plant needs in the Gulf States Utilities service area. The HTGR-PS/C was developed by General Atomic (GA) specifically for industries which require both steam and electric energy. The GA 1170-MW(t) HTGR-PC/C design is particularly well suited to industrial applications and is expected to have excellent cost benefits over other energy sources

  16. Mass-spectrometric determination in individual coated HTR fuel particles

    International Nuclear Information System (INIS)

    Strigl, A.

    1976-11-01

    A method is described which allows the simultaneous determination of fission and reaction gases in individual coated particles at temperatures up to 2000 0 C. The particles are heated under high-vacuum in a micro resistance-furnace up to the desired temperature. After preselected times the particles are crushed by action of a pneumatic cylinder. The gases liberated are fed into a quadrupoleanalyzer where they are analyzed in a dynamic mode. A peak selector allows the simultaneous measurement of up to four gases. The method is used routinely for the determination of fission gases (Kr and Xe) and of carbon monoxide which is formed as a reaction gas from oxide fuel. Precision and accuracy are in the order of a few percent. Detection limits for routine measurements are about 10 -7 cm 3 (STP) for KR and Xe and 2 x 10 -5 cm 3 (STP) for CO but can be lowered by special techniques. (author)

  17. Some calculations of the failure statistics of coated fuel particles

    International Nuclear Information System (INIS)

    Martin, D.G.; Hobbs, J.E.

    1977-03-01

    Statistical variations of coated fuel particle parameters were considered in stress model calculations and the resulting particle failure fraction versus burn-up evaluated. Variations in the following parameters were considered simultaneously: kernel diameter and porosity, thickness of the buffer, seal, silicon carbide and inner and outer pyrocarbon layers, which were all assumed to be normally distributed, and the silicon carbide fracture stress which was assumed to follow a Weibull distribution. Two methods, based respectively on random sampling and convolution of the variations were employed and applied to particles manufactured by Dragon Project and RFL Springfields. Convolution calculations proved the more satisfactory. In the present calculations variations in the silicon carbide fracture stress caused the greatest spread in burn-up for a given change in failure fraction; kernel porosity is the next most important parameter. (author)

  18. TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

    International Nuclear Information System (INIS)

    Smith, P.D.

    1978-02-01

    A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique

  19. HTGR Application Economic Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  20. TRISO-coated spent fuel processing using a Grind-Leach head-end

    International Nuclear Information System (INIS)

    Spencer, Barry B.; Del Cul, Guillermo D.; Mattus, Catherine H.; Collins, Emory D.

    2005-01-01

    Processing of TRISO-coated HTGR fuels with the grind-leach process requires that the fuel be finely pulverized for efficient and effective acid dissolution of the fuel components. Mechanical size reduction of the fuel is being investigated with jet mill technology as the final milling step. Laboratory experiments were performed with surrogates of crushed fuel compacts that indicate that milling to very small particle sizes is feasible. The size distribution of the milled product is sensitive to the solids feed rate, and the distribution may be bimodal which could support an effective solid-solid separation. (author)

  1. HTGR programme in the United States of America

    International Nuclear Information System (INIS)

    Fox, J.E.

    1991-01-01

    The HTGR is being developed by the US Department of Energy within the Division of HTGRs is reported. Fuel design, development and demonstration activities are being conducted by General Atomics and Oak Ridge National Laboratory. During FY-1990 the US continued work in cooperative projects with the KFA-Forschungszentrum Juelich and the Japan Atomic Energy Research Institute on post irradiation examination of fuel capsules and continued the Fission Product Transport Test Program with the French Commissariat a l'Energie Atomique in the COMEDIE in-pile loop at the SILOE reactor at Grenoble. Other activities included installation of the high temperature core-conduction-cooldown test furnace at ORNL which will be used for testing of irradiated fuel compacts under accident conditions. Finally, the US fuel performance experts participated in the MHTGR Cost Reduction Study which is a major effort within the US commercial MHTGR program. 1 tab

  2. Recent developments in graphite. [Use in HTGR and aerospace

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.

  3. HTGR Industrial Application Functional and Operational Requirements

    International Nuclear Information System (INIS)

    Demick, L.E.

    2010-01-01

    This document specifies the functional and performance requirements to be used in the development of the conceptual design of a high temperature gas-cooled reactor (HTGR) based plant supplying energy to a typical industrial facility. These requirements were developed from collaboration with industry and HTGR suppliers over the preceding three years to identify the energy needs of industrial processes for which the HTGR technology is technically and economically viable. The functional and performance requirements specified herein are an effective representation of the industrial sector energy needs and an effective basis for developing a conceptual design of the plant that will serve the broadest range of industrial applications.

  4. Aqueous alteration of VHTR fuels particles under simulated geological conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ait Chaou, Abdelouahed, E-mail: aitchaou@subatech.in2p3.fr; Abdelouas, Abdesselam; Karakurt, Gökhan; Grambow, Bernd

    2014-05-01

    Very High Temperature Reactor (VHTR) fuels consist of the bistructural-isotropic (BISO) or tristructural-isotropic (TRISO)-coated particles embedded in a graphite matrix. Management of the spent fuel generated during VHTR operation would most likely be through deep geological disposal. In this framework we investigated the alteration of BISO (with pyrolytic carbon) and TRISO (with SiC) particles under geological conditions simulated by temperatures of 50 and 90 °C and in the presence of synthetic groundwater. Solid state (scanning electron microscopy (SEM), micro-Raman spectroscopy, electron probe microanalyses (EPMA) and X-ray photoelectron spectroscopy (XPS)) and solution analyses (ICP-MS, ionique chromatography (IC)) showed oxidation of both pyrolytic carbon and SiC at 90 °C. Under air this led to the formation of SiO{sub 2} and a clay-like Mg–silicate, while under reducing conditions (H{sub 2}/N{sub 2} atmosphere) SiC and pyrolytic carbon were highly stable after a few months of alteration. At 50 °C, in the presence and absence of air, the alteration of the coatings was minor. In conclusion, due to their high stability in reducing conditions, HTR fuel disposal in reducing deep geological environments may constitute a viable solution for their long-term management.

  5. Support vector machine to predict diesel engine performance and emission parameters fueled with nano-particles additive to diesel fuel

    Science.gov (United States)

    Ghanbari, M.; Najafi, G.; Ghobadian, B.; Mamat, R.; Noor, M. M.; Moosavian, A.

    2015-12-01

    This paper studies the use of adaptive Support Vector Machine (SVM) to predict the performance parameters and exhaust emissions of a diesel engine operating on nanodiesel blended fuels. In order to predict the engine parameters, the whole experimental data were randomly divided into training and testing data. For SVM modelling, different values for radial basis function (RBF) kernel width and penalty parameters (C) were considered and the optimum values were then found. The results demonstrate that SVM is capable of predicting the diesel engine performance and emissions. In the experimental step, Carbon nano tubes (CNT) (40, 80 and 120 ppm) and nano silver particles (40, 80 and 120 ppm) with nanostructure were prepared and added as additive to the diesel fuel. Six cylinders, four-stroke diesel engine was fuelled with these new blended fuels and operated at different engine speeds. Experimental test results indicated the fact that adding nano particles to diesel fuel, increased diesel engine power and torque output. For nano-diesel it was found that the brake specific fuel consumption (bsfc) was decreased compared to the net diesel fuel. The results proved that with increase of nano particles concentrations (from 40 ppm to 120 ppm) in diesel fuel, CO2 emission increased. CO emission in diesel fuel with nano-particles was lower significantly compared to pure diesel fuel. UHC emission with silver nano-diesel blended fuel decreased while with fuels that contains CNT nano particles increased. The trend of NOx emission was inverse compared to the UHC emission. With adding nano particles to the blended fuels, NOx increased compared to the net diesel fuel. The tests revealed that silver & CNT nano particles can be used as additive in diesel fuel to improve complete combustion of the fuel and reduce the exhaust emissions significantly.

  6. An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

    International Nuclear Information System (INIS)

    Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

    2008-03-01

    In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

  7. On gas and particle radiation in pulverized fuel combustion furnaces

    DEFF Research Database (Denmark)

    Yin, Chungen

    2015-01-01

    Radiation is the principal mode of heat transfer in a combustor. This paper presents a refined weighted sum of gray gases model for computational fluid dynamics modelling of conventional air-fuel combustion, which has greater accuracy and completeness than the existing gaseous radiative property...... models. This paper also presents new conversion-dependent models for particle emissivity and scattering factor, instead of various constant values in literature. The impacts of the refined or new models are demonstrated via computational fluid dynamics simulation of a pulverized coal-fired utility boiler...

  8. A reactivity accidents simulation of the Fort Saint Vrain HTGR

    International Nuclear Information System (INIS)

    Fainer, Gerson

    1980-01-01

    A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

  9. Temperature of loose coated particles in irradiation tests

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1975-04-01

    An analysis is presented of the temperature of a monolayer bed of loose High-Temperature Gas-Cooled Reactor (HTGR) type fissioning fuel particles in an annular cavity. Both conduction and radiant heat transfer are taken into account, and the effect of particle contact with the annular cavity surfaces is evaluated. Charts are included for the determination of the maximum surface temperature of the particle coating for any size particle or power generation rate in a fuel bed of this type. The charts are intended for the design and evaluation of irradiation experiments on loose beds of coated fuel particles of the type used in HTGRs. Included in an Appendix is a method for estimating the temperature of a particle in circular hole. (U.S.)

  10. Proceedings of the 2nd JAERI symposium on HTGR technologies October 21 ∼ 23, 1992, Oarai, Japan

    International Nuclear Information System (INIS)

    1993-01-01

    The Japan Atomic Energy Research Institute (JAERI) held the 2nd JAERI Symposium on HTGR Technologies on October 21 to 23, 1992, at Oarai Park Hotel at Oarai-machi, Ibaraki-ken, Japan, with support of International Atomic Energy Agency (IAEA), Science and Technology Agency of Japan and the Atomic Energy Society of Japan on the occasion that the construction of the High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, is now being proceeded smoothly. In this symposium, the worldwide present status of research and development (R and D) of the HTGRs and the future perspectives of the HTGR development were discussed with 47 papers including 3 invited lectures, focusing on the present status of HTGR projects and perspectives of HTGR Development, Safety, Operation Experience, Fuel and Heat Utilization. A panel discussion was also organized on how the HTGRs can contribute to the preservation of global environment. About 280 participants attended the symposium from Japan, Bangladesh, Germany, France, Indonesia, People's Republic of China, Poland, Russia, Switzerland, United Kingdom, United States of America, Venezuela and the IAEA. This paper was edited as the proceedings of the 2nd JAERI Symposium on HTGR Technologies, collecting the 47 papers presented in the oral and poster sessions along with 11 panel exhibitions on the results of research and development associated to the HTTR. (author)

  11. Generator technology for HTGR power plants

    International Nuclear Information System (INIS)

    Lomba, D.; Thiot, D.

    1997-01-01

    Approximately 15% of the worlds installed capacity in electric energy production is from generators developed and manufactured by GEC Alsthom. GEC Alsthom is now working on the application of generators for HTGR power conversion systems. The main generator characteristics induced by the different HTGR power conversion technology include helium immersion, high helium pressure, brushless excitation system, magnetic bearings, vertical lineshaft, high reliability and long periods between maintenance. (author)

  12. Thermochemical equilibrium in a kernel of a UN TRISO coated fuel particle

    International Nuclear Information System (INIS)

    Kim, Young Min; Jo, C. K.; Lim, H. S.; Cho, M. S.; Lee, W. J.

    2012-01-01

    A coated fuel particle (CFP) with a uranium mononitride (UN) kernel has been recently considered as an advanced fuel option, such as in fully ceramic micro encapsulated (FCM) replacement fuel for light water reactors (LWRs). In FCM fuel, a large number of tri isotropic coated fuel particles (TRISOs) are embedded in a silicon carbide (SiC) matrix. Thermochemical equilibrium calculations can predict the chemical behaviors of a kernel in a TRISO of FCM fuel during irradiation. They give information on the kind and quantity of gases generated in a kernel during irradiation. This study treats the quantitative analysis of thermochemical equilibrium in a UN TRISO of FCM LWR fuel using HSC software

  13. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

    The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

  14. Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

    International Nuclear Information System (INIS)

    Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

    1977-12-01

    Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)

  15. Production control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.

    1979-06-01

    Purpose of this report was (1) to determine which techniques (Kalman Filter, weighted least squares, Shewhart control chart) are capable of detecting drift or step changes earliest in a manufacturing process, and (2) what method would work well in maintaining the manufacturing process at an acceptable level of quality. To solve part (1) simulation studies were performed for various test cases of interest. No single technique was superior in all of these cases, but the Kalman Filter appeared to be more robust to various process changes. The weighted least squares did a good job when the weight was near unity (0.9977) and failed when the weight was small (0.63). The Shewhart control chart is better for detecting step changes than for trends. Several methods wre compared to try to answer part (2). In this report the model building and forecasting was done using the methods of Box and Jenkins

  16. HTGR development in the United States of America

    International Nuclear Information System (INIS)

    Fox, J.E.

    1991-01-01

    The status of high temperature gas-cooled reactors (HTGR) development in the United States of America is described, including the organizational structure for the development support, HTGR development programme, and plans for future activities in the field

  17. Comparison of predicted and measured fission product behaviour in the Fort St. Vrain HTGR during the first three cycles of operation

    International Nuclear Information System (INIS)

    Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

    1985-01-01

    The 330 MW(e) Fort St. Vrain (M) High Temperature Gas-Cooled Reactor (HTGR) is fueled with (Th,U)C 2 /ThC 2 TRISO-coated fuel particles contained in prismatic graphite fuel elements. Fission product release from the reactor core has been monitored during the first three cycles of operation. In order to assess the validity of the design methods used to predict fission product source terms for HTGRs, fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible, which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors, which is consistent with plateout probe measurements. (author)

  18. Potential of the HTGR hydrogen cogeneration system in Japan

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Mouri, Tomoaki; Kunitomi, Kazuhiko

    2007-01-01

    A high temperature gas cooled reactor (HTGR) is one of the next generation nuclear systems. The HTGR hydrogen cogeneration system can produce not only electricity but also hydrogen. Then it has a potential to supply massive low-cost hydrogen without greenhouse gas emission for the future hydrogen society. Japan Atomic Energy Agency (JAEA) has been carried out the design study of the HTGR hydrogen cogeneration system (GTHTR300C). The thermal power of the reactor is 600 MW. The hydrogen production plant utilizes 370 MW and can supply 52,000 m 3 /h (0.4 Bm 3 /y) of hydrogen. Present industrial hydrogen production capacity in Japan is about 18 Bm 3 /y and it will decrease by 15 Bm 3 /y in 2030 due to the aging facilities. On the other hand, the hydrogen demand for fuel cell vehicle (FCV) in 2030 is estimated at 15 Bm 3 /y at a maximum. Since the hydrogen supply may be short after 2030, the additional hydrogen should be produced by clean hydrogen process to reduce greenhouse gas emission. This hydrogen shortage is a potential market for the GTHTR300C. The hydrogen production cost of GTHTR300C is estimated at 20.5 JPY/Nm 3 which has an economic competitiveness against other industrial hydrogen production processes. 38 units of the GTHTR300C can supply a half of this shortage which accounts for the 33% of hydrogen demand for FCV in 2100. According to the increase of hydrogen demand, the GTHTR300C should be constructed after 2030. (author)

  19. Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

    International Nuclear Information System (INIS)

    Omori, R.

    2000-01-01

    We have continued theoretical and experimental studies on laser manipulation of nuclear fuel particles, such as UO 2 , PuO 2 and ThO 2 , In this paper, we investigate the applicability of the collection of MOX particles floating in air using radiation pressure of a laser light; some preliminary results are shown. This technique will be useful for removal and confinement of MOX particles being transported by air current or dispersed in a cell box. First, we propose two types of principles for collecting MOX particles. Second, we show some experimental results, Third, we show numerical results of radiation pressure exerted on submicrometer-sized UO 2 particles using Generalized Lorentz-Mie theory. Because optical constants of UO 2 are similar to those of MOX fuel particles, it seems that calculation results obtained hold for MOX fuel particles. 2. Principles of collecting MOX fuel particles using radiation pressure (authors)

  20. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

    2008-01-01

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

  1. IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Rouxelin, Pascal Nicolas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented by the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise

  2. Effects of fuel particle size distributions on neutron transport in stochastic media

    International Nuclear Information System (INIS)

    Liang, Chao; Pavlou, Andrew T.; Ji, Wei

    2014-01-01

    Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron

  3. HTGR-INTEGRATED COAL TO LIQUIDS PRODUCTION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Anastasia M Gandrik; Rick A Wood

    2010-10-01

    As part of the DOE’s Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to “shift” the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700°C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: • 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal

  4. HTGR-Integrated Coal To Liquids Production Analysis

    International Nuclear Information System (INIS)

    Gandrik, Anastasia M.; Wood, Rick A.

    2010-01-01

    As part of the DOE's Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to 'shift' the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700 C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: (1) 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal consumption by 66

  5. Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

    International Nuclear Information System (INIS)

    Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

    2002-01-01

    Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and German and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration, power per particle) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance. (author)

  6. European research and development on HTGR process heat applications

    International Nuclear Information System (INIS)

    Verfondern, Karl; Lensa, Werner von

    2003-01-01

    The High-Temperature Gas-Cooled Reactor represents a suitable and safe concept of a future nuclear power plant with the potential to produce process heat to be utilized in many industrial processes such as reforming of natural gas, coal gasification and liquefaction, heavy oil recovery to serve for the production of the storable commodities hydrogen or energy alcohols as future transportation fuels. The paper will include a description of the broad range of applications for HTGR process heat and describe the results of the German long-term projects ''Prototype Nuclear Process Heat Reactor Project'' (PNP), in which the technical feasibility of an HTGR in combination with a chemical facility for coal gasification processes has been proven, and ''Nuclear Long-Distance Energy Transportation'' (NFE), which was the demonstration and verification of the closed-cycle, long-distance energy transmission system EVA/ADAM. Furthermore, new European research initiatives are shortly described. A particular concern is the safety of a combined nuclear/chemical facility requiring a concept against potential fire and explosion hazards. (author)

  7. THE INFLUENCE OF CARBON BURNOUT ON SUBMICRON PARTICLE FORMATION FROM EMULSIFIED FUEL OIL COMBUSTION

    Science.gov (United States)

    The paper gives results of an examination of particle behavior and particle size distributions from the combustion of different fuel oils and emulsified fuels in three experimental combusators. Results indicate that improved carbon (C) burnout from fule oil combustion, either by...

  8. 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.

  9. 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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  12. The prospects of HTGR in China

    International Nuclear Information System (INIS)

    Sun, Y.; Tong, Y.; Wu, Z.

    1994-01-01

    Present situations of the energy market in China are briefly introduced, while the forecast of the possible development of the Chinese energy market is shortly discussed. The discussion focuses on the expected roles of high temperature gas-cooled reactors (HTGR) in the Chinese energy market in the next century. The history and present status of the development of HTGR technologies in China are presented. In the National High-Tech Programme, a 10 MW helium-cooled test reactor (HTR-10) is projected to be built within this century. The main technical and safety features of the HTR-10 reactor are discussed. (author)

  13. Effect study of multi-bubbles on stress distribution of fuel particle

    International Nuclear Information System (INIS)

    Zhao Yi; Wang Xiaomin; Long Chongsheng

    2015-01-01

    The finite element model was proposed to simulate the process of the UO_2 dispersion fuel particle sustaining the internal pressure of multi-bubbles, and the stress distribution of fuel particle with intra-bubbles was calculated. The results show that when the bubbles line equidistantly along x axis, the max normal stress along y axis increases with the number of bubbles, meanwhile, the increment of the normal stress gradually decreases. There is a limit that the effect of bubble's number imposes on the max normal stress in the fuel particle. When multi-column of bubbles exist, the max normal stress along x axis in the fuel particle increases, and the max normal stress along y axis decreases with the increase of the number of bubble column. The stress concentration in the fuel particle decreases with the spacing radius ratio increasing. (authors)

  14. Moisture desorption in mechanically masticated fuels: effects of particle fracturing and fuelbed compaction

    Science.gov (United States)

    Jesse K. Kreye; J.Morgan Varner; Eric E. Knapp

    2012-01-01

    Mechanical mastication is increasingly used as a wildland fuel treatment, reducing standing trees and shrubs to compacted fuelbeds of fractured woody fuels. One major shortcoming in our understanding of these fuelbeds is how particle fracturing influences moisture gain or loss, a primary determinant of fire behaviour. To better understand fuel moisture dynamics, we...

  15. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  16. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

  17. Consolidated fuel reprocessing program. Progress report, January 1-March 31, 1981

    International Nuclear Information System (INIS)

    1981-06-01

    Progress and activities are reported on process development, laboratory R and D, engineering research, engineering systems, Integrated Equipment Test (IET) facility operations, and HTGR fuel reprocessing

  18. SP-100 coated-particle fuel development. Phase I. Final report

    International Nuclear Information System (INIS)

    1983-03-01

    This document is the final report of Phase I of the SP-100 Coated-Particle Fuel Development Program conducted by GA Technologies Inc. for the US Department of Energy under contract DE-AT03-82SF11690. The general objective of the study conducted between September and December 1982 was to evaluate coated-particle type fuel as an alternate or backup fuel to the UO 2 tile-and-fin arrangement currently incorporated into the reference design of the SP-100 reactor core. This report presents and discusses the following topics in the order listed: the need for an alternative fuel for the SP-100 nuclear reactor; an abbreviated description of the reference and coated-particle fuel module concepts; the bases and results of the study and analysis leading to the preliminary design of a coated particle suitable for the SP-100 space power reactor; incorporation of the fuel particles into compacts and heat-pipe-cooled modules; initial efforts and plans to fabricate coated-particle fuel and fuel compacts; the design and performance of the proposed alternative core relative that of the reference fuel; and a summary of critical issues and conclusions consistent with the level of effort and duration of the study

  19. High-quality thorium TRISO fuel performance in HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

    2013-11-01

    Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

  20. High-quality thorium TRISO fuel performance in HTGRs

    International Nuclear Information System (INIS)

    Verfondern, Karl; Allelein, Hans-Josef; Nabielek, Heinz; Kania, Michael J.

    2013-01-01

    Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O 2 TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10 25 n/m 2 (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10 -5 for manufacturing, 2 x 10 -4 for normal operating conditions, and 5 x 10 -4 for accident conditions. These

  1. Improvements in the preparation of nuclear fuel elements with addition of a molding mixture to fuel particles

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1975-01-01

    An improved molting mixture to be added to nuclear fuel particles for the preparation of nuclear fuel elements is presented. It consists of carbon and pitch particles and contains an additive reducing the final coke yield of the fuel mass formed. This additive is chosen from: polystyrene and copolymers of styrene and butadiene of molecular weight between 500 and 1000000; aromatic compounds of molecular weight between 75 and 300; saturated hydrocarbon polymers of molecular weight between 500 and 1000000. The additive may be camphor, naphthalene, anthracene, phenanthrene, dimethyl terephthalate or their mixtures and is present at a concentration of 5 to 50% by weight. The carbon particles used consist of powdered graphite. These fuel elements are intended for gas-cooled high-temperature reactors [fr

  2. TRISO coated fuel particles with enhanced SiC properties

    International Nuclear Information System (INIS)

    Lopez-Honorato, E.; Tan, J.; Meadows, P.J.; Marsh, G.; Xiao, P.

    2009-01-01

    The silicon carbide (SiC) layer used for the formation of TRISO coated fuel particles is normally produced at 1500-1650 deg. C via fluidized bed chemical vapor deposition from methyltrichlorosilane in a hydrogen environment. In this work, we show the deposition of SiC coatings with uniform grain size throughout the coating thickness, as opposed to standard coatings which have larger grain sizes in the outer sections of the coating. Furthermore, the use of argon as the fluidizing gas and propylene as a carbon precursor, in addition to hydrogen and methyltrichlorosilane, allowed the deposition of stoichiometric SiC coatings with refined microstructure at 1400 and 1300 deg. C. The deposition of SiC at lower deposition temperatures was also advantageous since the reduced heat treatment was not detrimental to the properties of the inner pyrolytic carbon which generally occurs when SiC is deposited at 1500 deg. C. The use of a chemical vapor deposition coater with four spouts allowed the deposition of uniform and spherical coatings.

  3. Method to manufacture spherical fuel and breeder particles

    International Nuclear Information System (INIS)

    Huschka, H.; Kadner, M.

    1976-01-01

    Optimum properties of the pyrolytic carbon cladding layer deposited on fuel and breeder cores are best achieved by forming the layers into exact spherical shells. It is necessary to have a uniform shperical shape of the cores to be coated. This is achieved by converting an oscillating liquid jet flowing out of one or several nozzles, of uranium and/or thorium solutions which drop into an ammonia solution at a quantity of over 3000 drops per minute. The drops prior to plunging into the ammonia solution, according to the invention, firstly run through an ammonia gasfree fall to acquire the shperical shape, then they fall through a zone flowed-through by ammonia gas. The ammonia gas is introduced into the dropping zone so that it flows in the opposite direction to falling and so that in addition a horizontal cross-flowing of the gas between the drops is guaranteed. The spherical drops are thus hardened before entering the ammonia solution. They are then washed as usual, dried and sintered. 4 examples are given to prepare thorium dioxide, uranium carbide and (U,Th) mixed oxide particles. (IHOE) [de

  4. Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

    International Nuclear Information System (INIS)

    Omori, R.; Suzuki, A.

    2001-01-01

    We proposed two principles based on the laser manipulation technique for collecting MOX fuel particles floating in air. While Principle A was based on the acceleration of the MOX particles due to the radiation pressure of a visible laser light, Principle B was based on the gradient forces exerted on the particles when an infrared laser light was incident. Principle A was experimentally verified using MnO 2 particles. Numerical results also showed the possibility of collecting MOX fuel particles based on both the principles. (authors)

  5. Regulatory Framework of Safety for HTGR

    International Nuclear Information System (INIS)

    Huh, Chang Wook; Suh, Nam Duk

    2011-01-01

    Recent accident in Fukushima Daiichi plant in Japan makes big impacts on the future of nuclear business. Many countries are changing their nuclear projects and increased safety of nuclear plants is asked for from the public. Without providing safety the society accepts, it might be almost impossible to build new plants further. In this sense high temperature gas-cooled reactor (HTGR) which is under development needs to be licensed reflecting this new expectation regarding safety. It means we should have higher level of safety goal and a systematic regulatory framework to assure the safety. In our previous paper, we evaluated the current safety goal and design practice in view of this new safety expectation after Fukushima accident. It was argued that a top-down approach starting from safety goal is necessary to develop safety requirements or to assure safety. Thus we need to propose an ultimate safety goal public accepts and then establish a systematic regulatory framework. In this paper we are going to provide a conceptual regulatory framework to guarantee the safety of HTGR. Section 2 discusses the recent trend of IAEA safety requirements and then summarize the HTGR design approach. Incorporating these discussions, we propose a conceptual framework of regulation for safety of HTGR

  6. FY1983 HTGR summary level program plan

    International Nuclear Information System (INIS)

    1983-01-01

    The major focus and priority of the FY1983 HTGR Program is the development of the HTGR-SC/C Lead Project through one of the candidate lead utilities. Accordingly, high priority will be given to work described in WBS 04 for site and user specific studies toward the development of the Lead Project. Asessment of advanced HTGR systems will continue during FY1983 in accordance with the High Temperature Process Heat (HTPH) Concept Evaluation Plan. Within the context of that plan, the assessment of the monolithic HTPH concepts has been essentially completed in FY1982 and FY1983 activities and will be limited to documentation only. the major advanced HTGR systems efforts in FY1983 will be focused on the further definition of the Modular Reactor Systems concepts in both the reforming (MRS-R) and Steam Cycle/Cogeneration 9MRS-SC/C) configurations in WBS 41. The effort will concentrate upon key technical issues and trade studies oriented to reduction in expected cost and schedule duration. With regard to the latter, the most significant will be trade study addressing the degree of modularization of reactor plant structures. particular attention will be given to the confinement building which currently defines the critical path for construction

  7. HTGR gas turbine power plant preliminary design

    International Nuclear Information System (INIS)

    Koutz, S.L.; Krase, J.M.; Meyer, L.

    1973-01-01

    The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references

  8. Advances in Automated QA/QC for TRISO Fuel Particle Production

    International Nuclear Information System (INIS)

    Hockey, Ronald L.; Bond, Leonard J.; Batishko, Charles R.; Gray, Joseph N.; Saurwein, John J.; Lowden, Richard A.

    2004-01-01

    Fuel in most Generation IV reactor designs typically encompasses billions of the TRISO particles. Present day QA/QC methods, done manually and in many cases destructively, cannot economically test a statistically significant fraction of the large number of the individual fuel particles required. Fully automated inspection technologies are essential to economical TRISO fuel particle production. A combination of in-line nondestructive (NDE) measurements employing electromagnetic induction and digital optical imaging analysis is currently under investigation and preliminary data indicate the potential for meeting the demands of this application. To calibrate high-speed NDE methods, surrogate fuel particle samples are being coated with layers containing a wide array of defect types found to degrade fuel performance and these are being characterized via high-resolution CT and digital radiographic images

  9. Fuel particles in the Chernobyl cooling pond: current state and prediction for remediation options

    International Nuclear Information System (INIS)

    Bulgakov, A.; Konoplev, A.; Smith, J.; Laptev, G.; Voitsekhovich, O.

    2009-01-01

    During the coming years, a management and remediation strategy for the Chernobyl cooling pond (CP) will be implemented. Remediation options include a controlled reduction in surface water level of the cooling pond and stabilisation of exposed sediments. In terrestrial soils, fuel particles deposited during the Chernobyl accident have now almost completely disintegrated. However, in the CP sediments the majority of 90 Sr activity is still in the form of fuel particles. Due to the low dissolved oxygen concentration and high pH, dissolution of fuel particles in the CP sediments is significantly slower than in soils. After the planned cessation of water pumping from the Pripyat River to the Pond, significant areas of sediments will be drained and exposed to the air. This will significantly enhance the dissolution rate and, correspondingly, the mobility and bioavailability of radionuclides will increase with time. The rate of acidification of exposed bottom sediments was predicted on the basis of acidification of similar soils after liming. Using empirical equations relating the fuel particle dissolution rate to soil and sediment pH allowed prediction of fuel particle dissolution and 90 Sr mobilisation for different remediation scenarios. It is shown that in exposed sediments, fuel particles will be almost completely dissolved in 15-25 years, while in parts of the cooling pond which remain flooded, fuel particle dissolution will take about a century

  10. Analytical Dancoff factor evaluations for reactor designs loaded with TRISO particle fuel

    International Nuclear Information System (INIS)

    Ji, Wei; Liang, Chao; Pusateri, Elise N.

    2014-01-01

    Highlights: • The Dancoff factors for randomly distributed TRISO fuel particles are evaluated. • A new “dual-sphere” model is proposed to predict Dancoff factors. • The new model accurately accounts for the coating regions of fuel particles. • High accuracy is achieved over a broad range of design parameters. • The new model can be used to analyze reactors with double heterogeneity. - Abstract: A new mathematical model, the dual-sphere model, is proposed to analytically evaluate Dancoff factors of TRISO fuel kernels based on the chord method. The accurate evaluation of fuel kernel Dancoff factors is needed when one analyzes nuclear reactors loaded with TRISO particle fuel. In these reactor designs, fuel kernels are randomly distributed and shield each other, causing a shadowing effect. The Dancoff factor is a quantitative measure of this effect and is determined by the spatial distribution of fuel kernels. A TRISO fuel particle usually consists of four layers that form a coating region outside the fuel kernel. When fuel particles are loaded in the reactor, the spatial distribution of fuel kernels can be affected by the thickness of the coating region. Therefore, the coating region should be taken into account in the calculation of Dancoff factors. However, the previous model, the single-sphere model, assumes no coating regions in the Dancoff factor predictions. To address this model deficiency, the dual-sphere model is proposed by deriving a new chord length distribution function between two fuel kernels that explicitly accounts for coating regions. The new model is employed to derive analytical solutions of infinite medium, intra-fuel pebble and intra-fuel compact/pin Dancoff factors over a wide range of volume packing fractions of TRISO fuel particles, varying from 2% to 60%. Comparisons are made with the predictions from the single-sphere model and reference Monte Carlo simulations. A significant improvement of the accuracy, over the ranges of

  11. HTGR-steam cycle/cogeneration plant economic potential

    International Nuclear Information System (INIS)

    1981-05-01

    The cogeneration of heat and electricity provides the potential for improved fuel utilization and corresponding reductions in energy costs. In the evaluation of the cogeneration plant product costs, it is advantageous to develop joint-product cost curves for alternative cogeneration plant models. The advantages and incentives for cogeneration are then presented in a form most useful to evaluate the various energy options. The HTGR-Steam Cycle/Cogeneration (SC/C) system is envisioned to have strong cogeneration potential due to its high-quality steam capability, its perceived nuclear siting advantages, and its projected cost advantages relative to coal. The economic information presented is based upon capital costs developed during 1980 and the economic assumptions identified herein

  12. 131I release from a HTGR during the LOFC accident

    International Nuclear Information System (INIS)

    Foley, J.E.

    1975-03-01

    The time-dependent release of 131 I from both the core and the containment building of a high temperature gas-cooled (HTGR) reactor during the loss of forced coolant (LOFC) accident is studied. A simplified core release model is combined with a containment building release model so that the total amount of the isotope released to the environment can be calculated. The time-dependent release of 131 I from the core during the LOFC accident is primarily a function of the time-dependent core temperatures and the failed fuel release constants. The most important factor in calculating the amount of the isotope released to the environment is the total amount released into the containment building. (U.S.)

  13. Physical and chemical analysis of interaction between oxide fuel and pyrocarbon coating of coated particles

    International Nuclear Information System (INIS)

    Lyutikov, R.A.; Kromov, Yu.F.; Chernikov, A.S.

    1991-01-01

    In terms of the model proposed the equilibrium pressure of gases (CO, Kr, Xe) in pyrocarbon-coated uranium dioxide fuel particles has been calculated, as function of the initial composition of the fuel (O/U), the design features of the coated particles, the fuel temperature, and the burnup. The possibility of reducing gas pressure in the particles by alloying the kernels with uranium carbide, and increasing the kernel capacity for retention of solid fission products by alloying the uranium oxide with aluminum-silicates, has been investigated. (author)

  14. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-04-01

    In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure and a simultaneous loss of forced cool of the core

  15. Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

    2016-10-15

    Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

  16. Utilization of plutonium in HTGR and its actinide production

    International Nuclear Information System (INIS)

    Karin, S.; Brogli, R.; Lefler, W.; Nordheim, L.

    1976-01-01

    The HTGR is a potential plutonium consumer. In this function it would burn plutonium, produce electricity and the valuable fissile isotope U-233. The advantages of this concept are discussed but particular attention is given to the production and the destruction of the higher actinides due to the high burnup achievable in such a system. The presence of the strong resonances in the plutonium isotopes demanded an extension of the methods for evaluation of self-shielding factors, a different structure for broad groups, and the adaptation of the reactor codes to these changes. Specifications for coated plutonium particles were developed. Also procedures were determined to evaluate the alpha ray and neutron emission rates of the actinide nuclides. First cycle calculations were carried out to establish in detail the characteristics of the plutonium reactors and their results are given

  17. An introduction to our activities supporting HTGR developments in Japan

    International Nuclear Information System (INIS)

    An, S.; Hayashi, T.; Tsuchie, Y.

    1997-01-01

    On the view point the most important for the HTGR development promotion now in Japan is to have people know about HTGR, the Research Association of HTGR Plants(RAHP) has paid the best efforts for making an appealing report for the past two years. The outline of the report is described with an introduction of some basic experiments done on the passive decay heat removal as one of the activities carried out in a member of the association. (author)

  18. Reprocessing flowsheet and material balance for MEU spent fuel

    International Nuclear Information System (INIS)

    Abraham, L.

    1978-10-01

    In response to nonproliferation concerns, the high-temperature gas-cooled reactor (HTGR) Fuel Recycle Development Program is investigating the processing requirements for a denatured medium-enriched uranium--thorium (MEU/Th) fuel cycle. Prior work emphasized the processing requirements for a high-enriched uranium--thorium (HEU/Th) fuel cycle. This report presents reprocessing flowsheets for an HTGR/MEU fuel recycle base case. Material balance data have been calculated for reprocessing of spent MEU and recycle fuels in the HTGR Recycle Reference Facility (HRRF). Flowsheet and mass flow effects in MEU-cycle reprocessing are discussed in comparison with prior HEU-cycle flowsheets

  19. Very small HTGR nuclear power plant concepts for special terrestrial applications

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1983-01-01

    The role of the very small nuclear power plant, of a few megawatts capacity, is perceived to be for special applications where an energy source as required but the following prevail: 1) no indigenous fossil fuel source, in long transport distances that add substantially to the cost of oil, coal in gas, and 3) secure long-term power production for defense applications with freedom from fuel supply lines. A small High Temperature Gas-Cooled reactor (HTGR) plant could provide the total energy needs for 1) a military installation, 2) an island base of strategic significance, 3) an industrial community or 4) an urban area. The small HTGR is regarded as a fixed-base installation (as opposed to a mobile system). All of the major components would be factory fabricated and transported to the site where emphasis would be placed on minimizing the construction time. The very small HTGR plant, currently in an early stage of design definition, has the potential for meeting the unique needs of the small energy user in both the military and private sectors. The plant may find acceptance for specialized applications in the industrialized nations and to meet the energy needs of developing nations. Emphasis in the design has been placed on safety, simplicity and compactness

  20. Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2007-01-01

    High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

  1. Army Demonstration of Light Obscuration Particle Counters for Monitoring Aviation Fuel Contamination

    Science.gov (United States)

    2013-05-07

    Hydraulic industry has utilized this technology for decades and created a mature process •Hydraulic industry has developed recognized calibration ...Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18/16/13 14/10/7 Pamas/Parker/Particle Solutions 19/17/12 U.S. Army 19...17/14/13* Diesel Fuel World Wide Fuel Charter 4th 18/16/13 DEF (AUST) 5695B 18/16/13 Bosch/Cummins 18/16/13 Donaldson 22/21/18 14/13/11 12/9/6 P ll

  2. Beginning-of-life neutronic analysis of a 3000-MW(t) HTGR

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1975-12-01

    The results of a study of safety-related neutronic characteristics for the beginning-of-life core of a 3000-MW(t) High-Temperature Gas-Cooled Reactor are presented. Emphasis was placed on the temperature-dependent reactivity effects of fuel, moderator, control poisons, and fission products. Other neutronic characteristics studied were gross and local power distributions, neutron kinetics parameters, control rod and other material worths and worth distributions, and the reactivity worth of a selected hypothetical perturbation in the core configuration. The study was performed for the most part using discrete-ordinates transport theory codes and neutron cross sections that were interpolated from a four-parameter nine-group library supplied by the HTGR vendor. A few comparison calculations were also performed using nine-group data generated with an independent cross-section processing code system. Results from the study generally agree well with results reported by the HTGR vendor

  3. New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

    International Nuclear Information System (INIS)

    McDonald, C.F.; Sonn, D.L.

    1983-01-01

    Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR

  4. Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)

    International Nuclear Information System (INIS)

    Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro

    2011-12-01

    We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)

  5. Characteristics of SME biodiesel-fueled diesel particle emissions and the kinetics of oxidation.

    Science.gov (United States)

    Jung, Heejung; Kittelson, David B; Zachariah, Michael R

    2006-08-15

    Biodiesel is one of the most promising alternative diesel fuels. As diesel emission regulations have become more stringent, the diesel particulate filter (DPF) has become an essential part of the aftertreatment system. Knowledge of kinetics of exhaust particle oxidation for alternative diesel fuels is useful in estimating the change in regeneration behavior of a DPF with such fuels. This study examines the characteristics of diesel particulate emissions as well as kinetics of particle oxidation using a 1996 John Deere T04045TF250 off-highway engine and 100% soy methyl ester (SME) biodiesel (B100) as fuel. Compared to standard D2 fuel, this B100 reduced particle size, number, and volume in the accumulation mode where most of the particle mass is found. At 75% load, number decreased by 38%, DGN decreased from 80 to 62 nm, and volume decreased by 82%. Part of this decrease is likely associated with the fact that the particles were more easily oxidized. Arrhenius parameters for the biodiesel fuel showed a 2-3times greater frequency factor and approximately 6 times higher oxidation rate compared to regular diesel fuel in the range of 700-825 degrees C. The faster oxidation kinetics should facilitate regeneration when used with a DPF.

  6. Effect of fuel particles' size variations on multiplication factor in pebble-bed nuclear reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2005-01-01

    The pebble-bed reactor (Pbr) spherical fuel element consists of two radial zones: the inner zone, in which the fissile material in form of the so-called TRISO particles is uniformly dispersed in graphite matrix and the outer zone, a shell of pure graphite. A TRISO particle is composed of a fissile kernel (UO 2 ) and several layers of carbon composites. The effect of TRISO particles' size variations and distance between them on PBR multiplication factor is studied using MCNP code. Fuel element is modelled in approximation of a cubical unit cell with periodic boundary condition. The multiplication factor of the fuel element depends on the size of the TRISO particles due to resonance self-shielding effect and on the inter-particle distance due to inter-kernel shadowing. (author)

  7. Fuel particle coating data. [Detailed information on coating runs at Los Alamos Scientific Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Hollabaugh, C.M.; Wagner, P.; Wahman, L.A.; White, R.W.

    1977-01-01

    Development of coating on nuclear fuel particles for the High-Temperature Fuels Technology program at the Los Alamos Scientific Laboratory included process studies for low-density porous and high-density isotropic carbon coats, and for ZrC and ''alloy'' C/ZrC coats. This report documents the data generated by these studies.

  8. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  9. Safety criteria for advanced HTGR concepts

    International Nuclear Information System (INIS)

    Kroeger, W.

    1989-01-01

    It is commonly agreed that advanced HTGR concepts must be licensable, which means that they must fulfil existing regulatory requirements. Furthermore, it is necessary to improve their public acceptance and they must even be suitable for urban sites. Therefore, they should be 'safer' than existing plants, which mainly means with respect to low-frequency or beyond-design severe accidents. Last but not least, the realization of advanced HTGR would be easier if commonly shared safety principles could be stated ensuring this further increased level of safety internationally. These qualitative statements need to be cast into quantitative guidelines which can be used as a rationale for safety evaluation. This paper tries to describe the status reached and to stimulate international activities. (author). 12 refs, 4 figs, 3 tabs

  10. HTGR experience, programs, and future applications

    International Nuclear Information System (INIS)

    Moore, R.A.; Kantor, M.E.; Brey, H.L.; Olson, H.G.

    1982-01-01

    This paper reviews the current status of the programs for the development of high-temperature gas-cooled reactors (HTGRs) in the major industrial countries of the world. Existing demonstration plants and facilities are briefly described, and national programs for exploiting the unique high-temperature capabilities of the HTGR for commercial production of electricity and in process steam/heat application are discussed. (orig.)

  11. Exergy analysis of HTGR-GT

    International Nuclear Information System (INIS)

    Cao Jianhua; Wang Jie; Yang Xiaoyong; Yu Suyuan

    2005-01-01

    The High Temperature Gas-cooled Reactor (HTGR) coupled with gas turbine for high efficiency in electricity production is supposed to be one of the candidates for the future nuclear power plants. The HTGR gas turbine cycle is theoretically based on the Brayton cycle with recuperated, intercooled and precooled sub-processes. In this paper, an exergy analysis of the Brayton Cycle on HTGR is presented. The analyses were done for four typical reactor outlet temperatures and the exergy loss distribution and exergy loss ratio of each sub-process was quantified. The results show that more than a half of the exergy loss takes place in the reactor, while the low pressure compressor (LPC), the high pressure compressor (HPC) and the intercooler denoted by compress system together, play a much small role in the contribution of exergy losses. With the rise of the reactor outlet temperature, both the exergy loss and exergy loss ratio of the reactor can be greatly cut down, so is the total exergy loss of the cycle; while the exergy loss ratios of the recuperator and precooler have a small rise. The total exergy efficiency of the cycle is quite high (50% more or less). (authors)

  12. Problems of dosimetry and risk assessment associated with inhalation of fuel particles

    International Nuclear Information System (INIS)

    Repin, V.S.; Nechaev, S.Y.; Bondarenko, O.A.; Bykorez, A.I.; Kononenko, L.I.

    1995-01-01

    This work deals with the problems of dosimetry and risk assessment associated with inhalation of fuel particles. Radioactive emission parameters and potential for assessment of the lung cancer risk with inhalation penetration of hot particles are described. (O.L.). 10 refs., 9 figs., 1 tab

  13. Small PWR 'PFPWR50' using cermet fuel of Th-Pu particles

    International Nuclear Information System (INIS)

    Hirayama, Takashi; Shimazu, Yoichiro

    2009-01-01

    An innovative concept of PFPWR50 has been studied. The main feature of PFPWR50 has been to adopt TRISO coated fuel particles in a conventional PWR cladding. Coated fuel particle provides good confining ability of fission products. But it is pointed out that swelling of SiC layer at low temperature by irradiation has possibilities of degrading the integrity of coated fuel particle in the LWR environment. Thus, we examined the use of Cermet fuel replacing SiC layer to Zr metal or Zr compound. And the nuclear fuel has been used as fuel compact, which is configured to fix coated fuel particles in the matrix material to the shape of fuel pellet. In the previous study, graphite matrix is adopted as the matrix material. According to the burnup calculations of the several fuel concepts with those covering layers, we decide to use Zr layer embedded in Zr metal base or ZrC layer with graphite matrix. But carbon has the problem at low temperature by irradiation as well as SiC. Therefore, Zr covering layer and Zr metal base are finally selected. The other feature of PFPWR50 concept has been that the excess reactivity is suppressed during a cycle by initially loading burnable poison (gadolinia) in the fuels. In this study, a new loading pattern is determined by combining 7 types of assemblies in which the gadolinia concentration and the number of the fuel rods with gadolinia are different. This new core gives 6.7 equivalent full power years (EFPY) as the core life of a cycle. And the excess reactivity is suppressed to less than 2.0%Δk/k during the cycle. (author)

  14. Performance limits of coated particle fuel. Part I. The significance of empirical performance diagrams and mathematical models in fuel development and power reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Graham, L. W.; Hick, H.

    1973-06-15

    This report introduces a general survey of our present knowledge and understanding of coated particle fuel performance. It defines first the reference power reactor conditions and the reference coated particle design on which the survey is centred. It describes then the typical strategy which has been followed in coated particle fuel development by the Dragon Project R & D Branch. Finally it shows the priorities which have governed the time scale and scope of fuel development and of the present review.

  15. EVALUATION OF THE IMPACT OF OIL PRESENCE IN THE AVIATION FUEL ON PARTICLE SIZE DISTRIBUTION

    Directory of Open Access Journals (Sweden)

    Remigiusz JASIŃSKI

    2017-03-01

    Full Text Available Emissions from aircraft engines represent a highly complex and important issue, which is related to the risk to human health. Particles emitted in urban areas and in the vicinity of airports affect air quality and have a particularly negative impact on airport workers. The development of measurement techniques and the methodology for evaluating exhaust emissions have allowed for the elaboration of appropriate procedures for the certification of aircraft and the enhancement of existing standards. Particulate matter emissions depend, among other things, on the composition of the fuel used and its additives. Some aircraft engine designs require a fuel additive in the form of oil, which ensures the proper operation of the fuel supply system. This article presents the results of studies conducted on jet engines powered by clean aviation fuel and fuel with the addition of oil. The aim of the study was to evaluate the effect of the addition of oil on the size distribution and concentration of emitted particles. It was found that, for small values of thrust, oil additive increases the concentration of particles. With an increase in the thrust force, the reduction of particles concentration was recorded in the case of the engine powered by fuel with oil additive. There was no significant effect of oil additive on the size distribution of emitted particles.

  16. Stress Calculation of a TRISO Coated Particle Fuel by Using a Poisson's Ratio in Creep Condition

    International Nuclear Information System (INIS)

    Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Kim, W. K.

    2007-01-01

    KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) project since 2004, has been developing a performance analysis code for the TRISO coated particle fuel named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. KAERI, on the other hand, is developing an ABAQUS based finite element(FE) model to cover the non-linear behaviors of a coated particle fuel such as cracking or debonding of the TRISO coating layers. Using the ABAQUS based FE model, verification calculations were carried out for the IAEA CRP-6 benchmark problems involving creep, swelling, and pressure. However, in this model the Poisson's ratio for elastic solution was used for creep strain calculation. In this study, an improvement is made for the ABAQUS based finite element model by using the Poisson's ratio in creep condition for the calculation of the creep strain rate. As a direct input of the coefficient in a creep condition is impossible, a user subroutine for the ABAQUS solution is prepared in FORTRAN for use in the calculations of the creep strain of the coating layers in the radial and hoop directions of the spherical fuel. This paper shows the calculation results of a TRISO coated particle fuel subject to an irradiation condition assumed as in the Miller's publication in comparison with the results obtained from the old FE model used in the CRP-6 benchmark calculations

  17. Failure mechanisms in high temperature gas cooled reactor fuel particles

    International Nuclear Information System (INIS)

    Soo, P.; Uneberg, G.; Sabatini, R.L.; Schweitzer, D.G.

    1979-01-01

    BISO coated UO 2 and ThO 2 particles were heated to high temperatures to determine failure mechanisms during hypothetical loss of coolant scenarios. Rapid failure begins when the oxides are reduced to liquid carbides. Several failure mechanisms are applicable, ranging from hole and crack formation in the coatings to catastrophic particle disintegration

  18. Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

  19. Pseudo three-dimensional modeling of particle-fuel packing using distinct element method

    International Nuclear Information System (INIS)

    Yuki, Daisuke; Takata, Takashi; Yamaguchi, Akira

    2007-01-01

    Vibration-based packing of sphere-pac fuel is a key technology in a nuclear fuel manufacturing. In the production process of sphere-pac fuel, a Mixed Oxide (MOX) fuel is formed to spherical form and is packed in a cladding tube by adding a vibration force. In the present study, we have developed a numerical simulation method to investigate the behavior of the particles in a vibrated tube using the Distinct Element Method (DEM). In general, the DEM requires a significant computational cost. Therefore we propose a new approach in which a small particle can move through the space between three larger particles even in the two-dimensional simulation. We take into account an equivalent three-dimensional effect in the equations of motion. Thus it is named pseudo three-dimensional modeling. (author)

  20. Experimental study of the form of "hot" steel particles on the ignition characteristics of liquid fuels

    Science.gov (United States)

    Zakharevich, Arkadiy V.

    2015-01-01

    The results of an experimental study of laws governing the ignition of liquid propellants (kerosene, diesel fuel and petroleum residue) by the single spherical steel particle heated to high temperatures are presented. Is carried out the comparison of the ignition delay times of the investigated flammable substances by the particles in the sphere and disk forms. It is established that the particle shape does not exert a substantial influence on the ignition process characteristics.

  1. Calculating failure probabilities for TRISO-coated fuel particles using an integral formulation

    International Nuclear Information System (INIS)

    Miller, Gregory K.; Maki, John T.; Knudson, Darrell L.; Petti, David A.

    2010-01-01

    The fundamental design for a gas-cooled reactor relies on the safe behavior of the coated particle fuel. The coating layers surrounding the fuel kernels in these spherical particles, termed the TRISO coating, act as a pressure vessel that retains fission products. The quality of the fuel is reflected in the number of particle failures that occur during reactor operation, where failed particles become a source for fission products that can then diffuse through the fuel element. The failure probability for any batch of particles, which has traditionally been calculated using the Monte Carlo method, depends on statistical variations in design parameters and on variations in the strengths of coating layers among particles in the batch. An alternative approach to calculating failure probabilities is developed herein that uses direct numerical integration of a failure probability integral. Because this is a multiple integral where the statistically varying parameters become integration variables, a fast numerical integration approach is also developed. In sample cases analyzed involving multiple failure mechanisms, results from the integration methods agree closely with Monte Carlo results. Additionally, the fast integration approach, particularly, is shown to significantly improve efficiency of failure probability calculations. These integration methods have been implemented in the PARFUME fuel performance code along with the Monte Carlo method, where each serves to verify accuracy of the others.

  2. Design and development on automated control system of coated fuel particle fabrication process

    International Nuclear Information System (INIS)

    Liu Malin; Shao Youlin; Liu Bing

    2013-01-01

    With the development trend of the large-scale production of the HTR coated fuel particles, the original manual control system can not meet the requirement and the automation control system of coated fuel particle fabrication in modern industrial grade is needed to develop. The comprehensive analysis aiming at successive 4-layer coating process of TRISO type coated fuel particles was carried out. It was found that the coating process could be divided into five subsystems and nine operating states. The establishment of DCS-type (distributed control system) of automation control system was proposed. According to the rigorous requirements of preparation process for coated particles, the design considerations of DCS were proposed, including the principle of coordinated control, safety and reliability, integration specification, practical and easy to use, and open and easy to update. A complete set of automation control system for coated fuel particle preparation process was manufactured based on fulfilling the requirements of these principles in manufacture practice. The automated control system was put into operation in the production of irradiated samples for HTRPM demonstration project. The experimental results prove that the system can achieve better control of coated fuel particle preparation process and meet the requirements of factory-scale production. (authors)

  3. Thermomechanical behavior of fuel particles in a matrix during reactor power excursions

    International Nuclear Information System (INIS)

    Brittan, R.O.; Smith, R.S.

    1977-01-01

    This work determines the largest particle size that can be used in fabricating fuel material without exceeding temperature or stress criteria during transient operation. To do this temperature distribution histories must be determined for various particle sizes and volume fractions using typical power densities histories of transient reactor operation. From these, the critical stresses are calculated. The model chosen to accomplish this is a spherical fuel particle in a spherical matrix shell. Heat flow and temperature continuity conditions are imposed at the interface, and a zero temperature gradient is specified at the outer radius of the matrix shell. The particle power density is assumed to be uniform radially. Provisions are made for uniform power density in the matrix to model gamma heating and power density in interface layers to allow for radiant and fission fragment heating. A computer code was prepared to solve the model performance, yielding the temperature and stress distribution histories. Material property variation with temperature is employed, along with a close mockup of the power density history during self-limiting reactor transients. To date, four fuel systems have been investigated: 1) UC.ZrC particles in graphite; 2) UO 2 particles in graphite; 3) UO 2 particles in chromium 4) UO 2 particles in stainless steel. The study indicates that the maximum allowable particle diameter varies as the square root of the initial transient period and of the particle volume fraction. The critical thermophysical parameter is the thermal diffusivity of the particle, since in all cases studied it is many times smaller than that of the matrix. That of the UC.ZrC solid solution particle is 5 or more times larger than that of the UO 2 particle. It was found that the particles of system 1) above could be about 4 times larger than that of the other sy

  4. Structures of the particles of the condensed dispersed phase in solid fuel combustion products plasma

    International Nuclear Information System (INIS)

    Samaryan, A.A.; Chernyshev, A.V.; Nefedov, A.P.; Petrov, O.F.; Fortov, V.E.; Mikhailov, Yu.M.; Mintsev, V.B.

    2000-01-01

    The results of experimental investigations of a type of dusty plasma which has been least studied--the plasma of solid fuel combustion products--were presented. Experiments to determine the parameters of the plasma of the combustion products of synthetic solid fuels with various compositions together with simultaneous diagnostics of the degree of ordering of the structures of the particles of the dispersed condensed phase were performed. The measurements showed that the charge composition of the plasma of the solid fuels combustion products depends strongly on the easily ionized alkali-metal impurities which are always present in synthetic fuel in one or another amount. An ordered arrangement of the particles of a condensed dispersed phase in structures that form in a boundary region between the high-temperature and condensation zones was observed for samples of aluminum-coated solid fuels with a low content of alkali-metal impurities

  5. Application of boron and gadolinium burnable poison particles in UO2 and PUO2 fuels in HTRs

    International Nuclear Information System (INIS)

    Kloosterman, J.L.

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble (fuel zone with radius of 2.5 cm surrounded with a 0.5 cm thick graphite layer) and burnable poison particles (BPPs) containing B 4 C made of pure 10 B or containing Gd 2 O 3 made of natural Gd. Two types of fuel were considered: UO 2 fuel made of 8% enriched uranium and PuO 2 fuel made of plutonium from LWR spent fuel. The radius of the BPP and the number of particles per fuel pebble were varied to find the flattest reactivity-to-time curve. For the UO 2 fuel, the reactivity swing is lowest (around 2%) for BPPs made of B 4 C with radius of 75 μm. In this case around 1070 BPPs per fuel pebble are needed. For the PuO 2 fuel to get a reactivity swing below 4%, the optimal radius of the BPP is the same, but the number of particles per fuel pebble should be around 1600. The optimal radius of the Gd 2 O 3 particles in the UO 2 fuel is about 10 times that of the B 4 C particles. The reactivity swing is around 3% when each fuel pebble contains only 9 BPPs with radius of 840 μm. The results of the Gd particles illustrate nicely the usage of black burnable poison particles introduced by Van Dam [Ann. Nuclear Energy 27 (2000) 733

  6. Nuclear fuel particles in the environment - characteristics, atmospheric transport and skin doses

    International Nuclear Information System (INIS)

    Poellaenen, R.

    2002-05-01

    In the present thesis, nuclear fuel particles are studied from the perspective of their characteristics, atmospheric transport and possible skin doses. These particles, often referred to as 'hot' particles, can be released into the environment, as has happened in past years, through human activities, incidents and accidents, such as the Chernobyl nuclear power plant accident in 1986. Nuclear fuel particles with a diameter of tens of micrometers, referred to here as large particles, may be hundreds of kilobecquerels in activity and even an individual particle may present a quantifiable health hazard. The detection of individual nuclear fuel particles in the environment, their isolation for subsequent analysis and their characterisation are complicated and require well-designed sampling and tailored analytical methods. In the present study, the need to develop particle analysis methods is highlighted. It is shown that complementary analytical techniques are necessary for proper characterisation of the particles. Methods routinely used for homogeneous samples may produce erroneous results if they are carelessly applied to radioactive particles. Large nuclear fuel particles are transported differently in the atmosphere compared with small particles or gaseous species. Thus, the trajectories of gaseous species are not necessarily appropriate for calculating the areas that may receive large particle fallout. A simplified model and a more advanced model based on the data on real weather conditions were applied in the case of the Chernobyl accident to calculate the transport of the particles of different sizes. The models were appropriate in characterising general transport properties but were not able to properly predict the transport of the particles with an aerodynamic diameter of tens of micrometers, detected at distances of hundreds of kilometres from the source, using only the current knowledge of the source term. Either the effective release height has been higher

  7. Nuclear fuel particles in the environment - characteristics, atmospheric transport and skin doses

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R

    2002-05-01

    In the present thesis, nuclear fuel particles are studied from the perspective of their characteristics, atmospheric transport and possible skin doses. These particles, often referred to as 'hot' particles, can be released into the environment, as has happened in past years, through human activities, incidents and accidents, such as the Chernobyl nuclear power plant accident in 1986. Nuclear fuel particles with a diameter of tens of micrometers, referred to here as large particles, may be hundreds of kilobecquerels in activity and even an individual particle may present a quantifiable health hazard. The detection of individual nuclear fuel particles in the environment, their isolation for subsequent analysis and their characterisation are complicated and require well-designed sampling and tailored analytical methods. In the present study, the need to develop particle analysis methods is highlighted. It is shown that complementary analytical techniques are necessary for proper characterisation of the particles. Methods routinely used for homogeneous samples may produce erroneous results if they are carelessly applied to radioactive particles. Large nuclear fuel particles are transported differently in the atmosphere compared with small particles or gaseous species. Thus, the trajectories of gaseous species are not necessarily appropriate for calculating the areas that may receive large particle fallout. A simplified model and a more advanced model based on the data on real weather conditions were applied in the case of the Chernobyl accident to calculate the transport of the particles of different sizes. The models were appropriate in characterising general transport properties but were not able to properly predict the transport of the particles with an aerodynamic diameter of tens of micrometers, detected at distances of hundreds of kilometres from the source, using only the current knowledge of the source term. Either the effective release height has

  8. Mixing fuel particles for space combustion research using acoustics

    Science.gov (United States)

    Burns, Robert J.; Johnson, Jerome A.; Klimek, Robert B.

    1988-01-01

    Part of the microgravity science to be conducted aboard the Shuttle (STS) involves combustion using solids, particles, and liquid droplets. The central experimental facts needed for characterization of premixed quiescent particle cloud flames cannot be adequately established by normal gravity studies alone. The experimental results to date of acoustically mixing a prototypical particulate, lycopodium, in a 5 cm diameter by 75 cm long flame tube aboard a Learjet aircraft flying a 20-sec low-gravity trajectory are described. Photographic and light detector instrumentation combine to measure and characterize particle cloud uniformity.

  9. Conceptual design of small-sized HTGR system (1). Major specifications and system designs

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Yan, Xing L.; Tachibana, Yukio

    2011-06-01

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2030s. The design philosophy is that the HTR50S is a high advanced reactor, which is reducing the R and D risk based on the HTTR design, upgrading the performance and reducing the cost for commercialization by utilizing the knowledge obtained by the HTTR operation and the GTHTR300 design. The major specifications of the HTR50S were determined and targets of the technology demonstration using the HTR50S (e.g., the increasing the power density, reduction of the number of uranium enrichment in the fuel, increasing the burn up, side-by-side arrangement between the reactor pressure vessel and the steam generator) were identified. In addition, the system design of HTR50S, which offers the capability of electricity generation, cogeneration of electricity and steam for a district heating and industries, was performed. Furthermore, a market size of small-sized HTGR systems was investigated. (author)

  10. Automatic X-ray inspection for escaped coated particles in spherical fuel elements of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yang, Min; Liu, Qi; Zhao, Hongsheng; Li, Ziqiang; Liu, Bing; Li, Xingdong; Meng, Fanyong

    2014-01-01

    As a core unit of HTGRs (high-temperature gas-cooled reactors), the quality of spherical fuel elements is directly related to the safety and reliability of HTGRs. In line with the design and performance requirements of the spherical fuel elements, no coated fuel particles are permitted to enter the fuel-free zone of a spherical fuel element. For fast and accurate detection of escaped coated fuel particles, X-ray DR (digital radiography) imaging with a step-by-step circular scanning trajectory was adopted for Chinese 10 MW HTGRs. The scanning parameters dominating the volume of the blind zones were optimized to ensure the missing detection of the escaped coated fuel particles is as low as possible. We proposed a dynamic calibration method for tracking the projection of the fuel-free zone accurately, instead of using a fuel-free zone mask of fixed size and position. After the projection data in the fuel-free zone were extracted, image and graphic processing methods were combined for automatic recognition of escaped coated fuel particles, and some practical inspection results were presented. - Highlights: • An X-ray DR imaging system for quality inspection of spherical fuel elements was introduced. • A method for optimizing the blind-zone-related scanning parameter was proposed. • A dynamic calibration method for tracking the fuel-free zone accurately was proposed. • Some inspection results of the disqualified spherical fuel elements with escaped coated fuel particles were presented

  11. Fuel Property, Emission Test, and Operability Results from a Fleet of Class 6 Vehicles Operating on Gas-to-Liquid Fuel and Catalyzed Diesel Particle Filters

    Energy Technology Data Exchange (ETDEWEB)

    Alleman, T. L.; Eudy, L.; Miyasato, M.; Oshinuga, A.; Allison, S.; Corcoran, T.; Chatterjee, S.; Jacobs, T.; Cherrillo, R. A.; Clark, R.; Virrels, I.; Nine, R.; Wayne, S.; Lansing, R.

    2005-11-01

    A fleet of six 2001 International Class 6 trucks operating in southern California was selected for an operability and emissions study using gas-to-liquid (GTL) fuel and catalyzed diesel particle filters (CDPF). Three vehicles were fueled with CARB specification diesel fuel and no emission control devices (current technology), and three vehicles were fueled with GTL fuel and retrofit with Johnson Matthey's CCRT diesel particulate filter. No engine modifications were made.

  12. Adaptive neuro-fuzzy inference system (ANFIS) to predict CI engine parameters fueled with nano-particles additive to diesel fuel

    Science.gov (United States)

    Ghanbari, M.; Najafi, G.; Ghobadian, B.; Mamat, R.; Noor, M. M.; Moosavian, A.

    2015-12-01

    This paper studies the use of adaptive neuro-fuzzy inference system (ANFIS) to predict the performance parameters and exhaust emissions of a diesel engine operating on nanodiesel blended fuels. In order to predict the engine parameters, the whole experimental data were randomly divided into training and testing data. For ANFIS modelling, Gaussian curve membership function (gaussmf) and 200 training epochs (iteration) were found to be optimum choices for training process. The results demonstrate that ANFIS is capable of predicting the diesel engine performance and emissions. In the experimental step, Carbon nano tubes (CNT) (40, 80 and 120 ppm) and nano silver particles (40, 80 and 120 ppm) with nanostructure were prepared and added as additive to the diesel fuel. Six cylinders, four-stroke diesel engine was fuelled with these new blended fuels and operated at different engine speeds. Experimental test results indicated the fact that adding nano particles to diesel fuel, increased diesel engine power and torque output. For nano-diesel it was found that the brake specific fuel consumption (bsfc) was decreased compared to the net diesel fuel. The results proved that with increase of nano particles concentrations (from 40 ppm to 120 ppm) in diesel fuel, CO2 emission increased. CO emission in diesel fuel with nano-particles was lower significantly compared to pure diesel fuel. UHC emission with silver nano-diesel blended fuel decreased while with fuels that contains CNT nano particles increased. The trend of NOx emission was inverse compared to the UHC emission. With adding nano particles to the blended fuels, NOx increased compared to the net diesel fuel. The tests revealed that silver & CNT nano particles can be used as additive in diesel fuel to improve combustion of the fuel and reduce the exhaust emissions significantly.

  13. Research on in-pile release of fission products from coated particle fuels

    International Nuclear Information System (INIS)

    Fukuda, K.; Iwamoto, K.

    1985-01-01

    Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

  14. Particle confinement and fueling effects on the Maryland spheromak

    International Nuclear Information System (INIS)

    Filuk, A.B.

    1991-01-01

    The spheromak plasma confinement concept provides the opportunity to study the evolution of a nearly force-free magnetic field configuration. The plasma currents and magnetic fields are produced self-consistently, making this type of device attractive as a possible fusion reactor. At present, spheromaks are observed to have poorer particle and magnetic confinement than expected from simple theory. The purpose of this study is to examine the role of plasma density in the decay of spheromaks produced in the Maryland Spheromak experiment. Density measurements are made with an interferometer and Langmuir probe, and results are correlated with those of other plasma diagnostics to understand the sources of plasma, the spheromak formation effects on the density, and the magnitude of particle loss during the spheromak decay. A power and particle balance computer model is constructed and applied to the spheromaks studied in order to assess the impact of high density and particle loss rate on the spheromak decay. The observations and model indicate that the decay of the spheromaks is at present dominated by impurity radiation loss. The model also predicts that high density and short particle confinement time play a critical role in the spheromak power balance when the impurity levels are reduced

  15. Present status of HTGR research and development, 1995

    International Nuclear Information System (INIS)

    1996-02-01

    Based on the Long-term Program for Development and Utilization of Nuclear Energy which was revised in 1987, the Japan Atomic Energy Research Institute (JAERI) has carried out the Research and Development (R and D) on the High Temperature Gas-cooled Reactors (HTGRs) in Japan. The JAERI obtained the installation permit of the High Temperature Engineering Test Reactor (HTTR) from the Government in November 1990 and started the construction of the HTTR facility in the Oarai Research Establishment in March 1991. The HTTR is a test reactor with thermal output of 30MW and outlet coolant temperature of 850degC at the rated operation and 950degC at the high temperature test operation, using the pin-in-block type fuel, and has capability to demonstrate nuclear process heat utilization. The reactor pressure vessel and intermediate heat exchanger were installed in the reactor containment vessel in 1994, and reactor internals were also installed in the reactor pressure vessel in 1995. The first criticality will be attained in December 1997. This report describes the design outline and construction progress of the HTTR, R and D of fuel, materials and components for the HTGR and high temperature nuclear heat application, and innovative and basic researches for high temperature technologies at the HTTR. (J.P.N.)

  16. Characterization of airborne plutonium-bearing particles from a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Sanders, S.M. Jr.

    1977-11-01

    The elemental compositions, sizes, structures, and 239 Pu contents were determined for 299 plutonium-bearing particles isolated from airborne particles collected at various locations in the exhaust from a nuclear fuel reprocessing facility. These data were compared with data from natural aerosol particles. Most of the collected particles were composed of aggregates of crustal materials. Seven percent of the particles were organic and 3% were metallic, viz., iron, chromium, and nickel. High enrichment factors for titanium, manganese, chromium, nickel, zinc, and copper were evidence of the anthropic nature of some of the particles. The amount of plutonium in most particles was very small (less than one femtocurie of 239 Pu). Plutonium concentrations were determined by the fission track counting method. Only one particle contained sufficient plutonium for detection by electron microprobe analysis. This was a 1-μm-diameter particle containing 73% PuO 2 by weight (estimated to be 170 fCi of 239 Pu) in combination with Fe 2 O 3 and mica. The plutonium-bearing particles were generally larger than natural aerosols. The geometric mean diameter of those collected from the mechanical line exhaust point where plutonium is converted to the metal was larger than that of particles collected from the wet cabinet exhaust (13.7 μm vs. 4.6 μm). Particles from the mechanical line also contained more plutonium per particle than those from the wet cabinets

  17. Device for the separation of spherically shaped fuel or breeding material particles for nuclear reactors

    International Nuclear Information System (INIS)

    Gyarmati, E.; Muenzer, R.

    1974-01-01

    Spherical fuel or blanket material particles are graded by diameter. The particles, which are present in a loose pebble bed, are singulized by means of a drum and by pneumatic suction. Next they pass through a drop section past an optical barrier which generates pulses corresponding to the number of particles. The particles then run through an eccentric wheel. This generates an electric voltage across a potentiometer which corresponds to the size of the particles. The slider of the potentiometer is connected with the axle of the eccentric wheel whose distance to the wall of the drop canal varies between the largest and the smallest possible diameters of the particles over half a revolution. Another barrier downstream of the eccentric wheel causes the particles to be graded in different containers in accordance with their diameters determined in this way. (DG) [de

  18. Auxiliary plasma heating and fueling models for use in particle simulation codes

    International Nuclear Information System (INIS)

    Procassini, R.J.; Cohen, B.I.

    1989-01-01

    Computational models of a radiofrequency (RF) heating system and neutral-beam injector are presented. These physics packages, when incorporated into a particle simulation code allow one to simulate the auxiliary heating and fueling of fusion plasmas. The RF-heating package is based upon a quasilinear diffusion equation which describes the slow evolution of the heated particle distribution. The neutral-beam injector package models the charge exchange and impact ionization processes which transfer energy and particles from the beam to the background plasma. Particle simulations of an RF-heated and a neutral-beam-heated simple-mirror plasma are presented. 8 refs., 5 figs

  19. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  20. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  1. Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

    International Nuclear Information System (INIS)

    Benchrif, A.; Chetaine, A.; Amsil, H.

    2013-01-01

    Highlights: • AFPR-100 reactor considered as a small nuclear reactor without on-site refueling originally based on TRISO micro-fuel element. • The AFPR-100 reactor was re-designed using the new Spherical Cermet fuel element. • The adoption of the Cermet fuel instead of TRISO fuel reduces the core lifetime operation by 3.1 equivalent full power years. • We discussed the new micro-fuel element candidate for small and medium sized reactors. - Abstract: The Atoms For Peace Reactor (AFPR-100), as a 100 MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY

  2. Dynamic response of a multielement HTGR core

    International Nuclear Information System (INIS)

    Reich, M.; Bezler, P.; Koplik, B.; Curreri, J.; Goradia, H.; Lasker, L.

    1977-01-01

    One of the primary factors in determining the structural integrity and consequently the safety of a High Temperature Gas-Cooled Reactor (HTGR) is the dynamic response of the core when subjected to a seismic excitation. The HTGR core under consideration consists of several thousands of hexagonal elements arranged in vertical stacks containing about eight elements per stack. There are clearance gaps between adjacent elements, which can change substantially due to radiation effects produced during their active lifetime. Surrounding the outer periphery of the core are reflector blocks and restraining spring-pack arrangements which bear against the reactor vessel structure (PCRV). Earthquake input motions to this type of core arrangement will result in multiple impacts between adjacent elements as well as between the reflector blocks and the restraining spring packs. The highly complex nonlinear response associated with the multiple collisions across the clearance gaps and with the spring packs is the subject matter of this paper. Of particular importance is the ability to analyze a complex nonlinear system with gaps by employing a model with a reduced number of masses. This is necessary in order to obtain solutions in a time-frame and at a cost which is not too expensive. In addition the effect of variations in total clearance as well as the initial distribution of clearances between adjacent elements is of primary concern. Both of these aspects of the problem are treated in the present analysis. Finally, by constraining the motion of the reflector blocks, a more realistic description of the dynamic response of the multi-element HTGR core is obtained

  3. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    Energy Technology Data Exchange (ETDEWEB)

    Malevski, A L; Stoliarevski, A Ya; Vladimirov, V T; Larin, E A; Lesnykh, V V; Naumov, Yu V; Fedotov, I L

    1990-07-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

  4. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    International Nuclear Information System (INIS)

    Malevski, A.L.; Stoliarevski, A.Ya.; Vladimirov, V.T.; Larin, E.A.; Lesnykh, V.V.; Naumov, Yu.V.; Fedotov, I.L.

    1990-01-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

  5. Coated Particle and Deep Burn Fuels Monthly Highlights December 2010

    International Nuclear Information System (INIS)

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

    2011-01-01

    During FY 2011 the CP and DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for November 2010, ORNL/TM-2010/323, was distributed to program participants on December 9, 2010. The final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Pebble Bed Design (INL), (c) Radiation Damage and Properties; (2) TRISO (tri-structural isotropic) Development - (a) TRU (transuranic elements) Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing (ORNL); (4) Fuel Performance and Analytical Analysis - Fuel Performance Modeling (ORNL).

  6. Quantitative HTGR safety and forced outage goals

    International Nuclear Information System (INIS)

    Houghton, W.J.; Parme, L.L.; Silady, F.A.

    1985-05-01

    A key step in the successful implementation of the integrated approach is the definition of the overall plant-level goals. To be effective, the goals should provide clear statements of what is to be achieved by the plant. This can be contrasted to the current practice of providing design-prescriptive criteria which implicitly address some higher-level objective but restrict the designer's flexibility. Furthermore, the goals should be quantifiable in such a way that satisfaction of the goal can be measured. In the discussion presented, two such plant-level goals adopted for the HTGR and addressing the impact of unscheduled occurrences are described. 1 fig

  7. Selection of JAERI'S HTGR-GT concept

    International Nuclear Information System (INIS)

    Muto, Y.; Ishiyama, S.; Shiozawa, S.

    2001-01-01

    In JAERI, a feasibility study of HTGR-GT has been conducted as an assigned work from STA in Japan since January 1996. So far, the conceptual or preliminary designs of 600, 400 and 300 MW(t) power plants have been completed. The block type core and pebble-bed core have been selected in 600 MW(t) and 400/300 MW(t), respectively. The gas-turbine system adopts a horizontal single shaft rotor and then the power conversion vessel is separated into a turbine vessel and a heat exchanger vessel. In this paper, the issues related to the selection of these concepts are technically discussed. (author)

  8. Postirradiation examination and evaluation of Fort St. Vrain fuel element 1-0743

    International Nuclear Information System (INIS)

    Saurwein, J.J.; Miller, C.M.; Young, C.A.

    1981-05-01

    Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680 0 C, fissile and fertile particle burnups of approximately 6.2% and 0.3%, respectively, and a total burnup of 12,210 MWd/tonne. The postirradiation examination revealed that the element was in excellent condition. No cracks were observed on any of the element surfaces. The structural integrity of the fuel rods was good. No evidence of mechanical interaction between the fuel rods and fuel body was observed. Calculated irradiation parameters obtained with HTGR design codes were compared with measured data. Radial and axial power distributions, irradiation temperatures, neutron fluences, and fuel burnups were in good agreement with measurements. Calculated fuel rod strains were about a factor of three greater than were observed

  9. HTGR Application Economic Model Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  10. Particle and NO{sub x} Emissions from a HVO-Fueled Diesel Engine

    Energy Technology Data Exchange (ETDEWEB)

    Happonen, M.

    2012-10-15

    Concerns about oil price, the strengthening climate change and traffic related health effects are all reasons which have promoted the research of renewable fuels. One renewable fuel candidate is diesel consisting of hydrotreated vegetable oils (HVO). The fuel is essentially paraffinic, has high cetane number (>80) and contains practically no oxygen, aromatics or sulphur. Furthermore, HVO fuel can be produced from various feedstocks including palm, soybean and rapeseed oils as well as animal fats. HVO has also been observed to reduce all regulated engine exhaust emissions compared to conventional diesel fuel. In this thesis, the effect of HVO fuel on engine exhaust emissions has been studied further. The thesis is roughly divided into two parts. The first part explores the emission reductions associated with the fuel and studies techniques which could be applied to achieve further emission reductions. One of the studied techniques was adjusting engine settings to better suit HVO fuel. The settings chosen for adjustments were injection pressure, injection timing, the amount of EGR and the timing of inlet valve closing (with constant inlet air mass flow, i.e. Miller timing). The engine adjustments were also successfully targeted to reduce either NO{sub x} or particulate emissions or both. The other applied emission reduction technique was the addition of oxygenate to HVO fuel. The chosen oxygenate was di-n-pentyl ether (DNPE), and tested fuel blend included 20 wt-% DNPE and 80 wt-% HVO. Thus, the oxygen content of the resulting blend was 2 wt-%. Reductions of over 25 % were observed in particulate emissions with the blend compared to pure HVO while NOx emissions altered under 5 %. On the second part of this thesis, the effect of the studied fuels on chosen surface properties of exhaust particles were studied using tandem differential mobility analyzer (TDMA) techniques and transmission electron microscopy (TEM). The studied surface properties were oxidizability and

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

    Science.gov (United States)

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

    2013-06-01

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

  12. Neutronics feasibility of using Gd2O3 particles in VVER-1000 fuel assembly

    International Nuclear Information System (INIS)

    Hoang Van Khanh; Hoang Thanh Phi Hung; Tran Hoai Nam

    2016-01-01

    Neutronics feasibility of using Gd 2 O 3 particles for controlling excess reactivity of VVER-1000 fuel assembly has been investigated. The motivation is that the use of Gd 2 O 3 particles would increase the thermal conductivity of the UO 2 +Gd 2 O 3 fuel pellet which is one of the desirable characteristics for designing future high burnup fuel. The calculation results show that the Gd 2 O 3 particles with the diameter of 60 µm could control the reactivity similarly to that of homogeneous mixture with the same amount of Gd 2 O 3 . The power densities at the fuel pin with Gd 2 O 3 particles increase by about 10-11%, leading to the decrease of the power peak and a slightly flatter power distribution. The power peak appears at the periphery pins at the beginning of burnup process which is decreased by 0.9 % when using Gd 2 O 3 particles. Further work and improvement are being planned to optimize the high power peaking at the beginning of burnup. (author)

  13. Chlorine gas processing of oxide nuclear fuel particles containing thorium

    International Nuclear Information System (INIS)

    Knotik, K.; Bildstein, H.; Falta, G.; Wagner, H.

    Experimental studies on the chloride extraction and separation of U and Th from coated Th--U oxide particles are reported. After a description of the chlorination equipment and the experimental procedures, the results are discussed. The yield of U is determined as a function of the reaction temperature. The results of a thermogravimetric analysis of the chlorination of uranium carbide and thorium carbides are reported and used to establish the reaction mechanism for the chlorination

  14. Measurement of particle size distribution and mass concentration of nuclear fuel aerosols

    International Nuclear Information System (INIS)

    Pickering, S.

    1982-01-01

    The particle size distribution and particle mass concentration of a nuclear fuel aerosol is measured by admitting the aerosol into a vertically-extending container, positioning an alpha particle detector within the container so that its window is horizontal and directed vertically, stopping the admission of aerosol into the container, detecting the alpha-activity of the particles of the aerosol sedimenting onto the detector window (for example in a series of equal time intervals until a constant level is reached), and converting the alpha-activity measurements into particle size distribution and/or particle mass concentration measurements. The detector is attached to a pivotted arm and by raising a counterweight can be lowered from the container for cleaning. (author)

  15. Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Yoon-Sang; Lee, Don-Bae; Sohn, Woong-Hee; Hong, Soon-Hyung

    1998-01-01

    A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U 3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)

  16. User's manual for the Composite HTGR Analysis Program (CHAP-1)

    International Nuclear Information System (INIS)

    Gilbert, J.S.; Secker, P.A. Jr.; Vigil, J.C.; Wecksung, M.J.; Willcutt, G.J.E. Jr.

    1977-03-01

    CHAP-1 is the first release version of an HTGR overall plant simulation program with both steady-state and transient solution capabilities. It consists of a model-independent systems analysis program and a collection of linked modules, each representing one or more components of the HTGR plant. Detailed instructions on the operation of the code and detailed descriptions of the HTGR model are provided. Information is also provided to allow the user to easily incorporate additional component modules, to modify or replace existing modules, or to incorporate a completely new simulation model into the CHAP systems analysis framework

  17. 1170-MW(t) HTGR-PS/C plant application study report: SRC-II process application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    The solvent refined coal (SRC-II) process is an advanced process being developed by Gulf Mineral Resources Ltd. (a Gulf Oil Corporation subsidiary) to produce a clean, non-polluting liquid fuel from high-sulfur bituminous coals. The SRC-II commercial plant will process about 24,300 tonnes (26,800 tons) of feed coal per stream day, producing primarily fuel oil plus secondary fuel gases. This summary report describes the integration of a high-temperature gas-cooled reactor operating in a process steam/cogeneration mode (HTGR-PS/C) to provide the energy requirements for the SRC-II process. The HTGR-PS/C plant was developed by General Atomic Company (GA) specifically for industries which require energy in the form of both steam and electricity. General Atomic has developed an 1170-MW(t) HTGR-PS/C design which is particularly well suited to industrial applications and is expected to have excellent cost benefits over other sources of energy

  18. Contributions of fuel combustion to pollution by airborne particles in urban and non-urban environments

    International Nuclear Information System (INIS)

    1995-06-01

    The application of ion beam analysis (IBA) techniques to aerosol pollution problems has been used in a number of countries since the late 1970's and early 1980's. The technique, however, had not been tested in Australia. This document is the final report of a project which aimed to establish a fine particle monitoring network covering the greater Wollongong/Sydney/ Newcastle ares, investigate the relationships between fuel combustion and fine particle aerosols in urban and non urban environments, add to the limited database of baseline information on concentrations of fine particles resulting from such processes as fossil fuel burning and industrial manufacturing, identify and quantify sources of fine particles in New South Wales, and introduce into Australia accelerator based IBA techniques for the analysis of filter papers obtained from large scale monitoring networks. These objectives were addressed by the project which identified and quantified some sources of fine particles and established some relationships between fuel combustion and fine aerosols. More work is required to fully quantify relationships between natural and anthropogenic fine particle sources. 24 tabs., 44 figs., 83 refs

  19. Pressure analysis in the fabrication process of TRISO UO2-coated fuel particle

    International Nuclear Information System (INIS)

    Liu Malin; Shao Youlin; Liu Bing

    2012-01-01

    Highlights: ► The pressure signals during the real TRISO UO2-coated fuel particle fabrication process. ► A new relationship about the pressure drop change and the coated fuel particles properties. ► The proposed relationship is validated by experimental results during successive coating. ► A convenient method for monitoring the fluidized state during coating process. - Abstract: The pressure signals in the coating furnace are obtained experimentally from the TRISO UO 2 -coated fuel particle fabrication process. The pressure signals during the coating process are analyzed and a simplified relationship about the pressure drop change due to the coated layer is proposed based on the spouted bed hydrodynamics. The change of pressure drop is found to be consistent with the change of the combination factor about particle density, bed density, particle diameter and static bed height, during the successive coating process of the buffer PyC, IPyC, SiC and OPyC layer. The newly proposed relationship is validated by the experimental values. Based on this relationship, a convenient method is proposed for real-time monitoring the fluidized state of the particles in a high-temperature coating process in the spouted bed. It can be found that the pressure signals analysis is an effective method to monitor the fluidized state on-line in the coating process at high temperature up to 1600 °C.

  20. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

  1. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    International Nuclear Information System (INIS)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi

    2016-01-01

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible

  2. A comparative study of the number and mass of fine particles emitted with diesel fuel and marine gas oil (MGO)

    Science.gov (United States)

    Nabi, Md. Nurun; Brown, Richard J.; Ristovski, Zoran; Hustad, Johan Einar

    2012-09-01

    The current investigation reports on diesel particulate matter emissions, with special interest in fine particles from the combustion of two base fuels. The base fuels selected were diesel fuel and marine gas oil (MGO). The experiments were conducted with a four-stroke, six-cylinder, direct injection diesel engine. The results showed that the fine particle number emissions measured by both SMPS and ELPI were higher with MGO compared to diesel fuel. It was observed that the fine particle number emissions with the two base fuels were quantitatively different but qualitatively similar. The gravimetric (mass basis) measurement also showed higher total particulate matter (TPM) emissions with the MGO. The smoke emissions, which were part of TPM, were also higher for the MGO. No significant changes in the mass flow rate of fuel and the brake-specific fuel consumption (BSFC) were observed between the two base fuels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-06-15

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

  4. 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

  5. Determination of uranium in coated fuel particle compact by potassium fluoride fusion-gravimetric method

    International Nuclear Information System (INIS)

    Ito, Mitsuo; Iso, Shuichi; Hoshino, Akira; Suzuki, Shuichi.

    1992-03-01

    Potassium fluoride-gravimetric method has been developed for the determination of uranium in TRISO type-coated fuel particle compact. Graphite matrix in the fuel compact is burned off by heating it in a platinum crucible at 850degC. The coated fuel particles thus obtained are decomposed by fusion with potassium fluoride at 900degC. The melt was dissolved with sulfuric acid. Uranium is precipitated as ammonium diuranate, by passing ammonia gas through the solution. The resulting precipitate is heated in a muffle furnace at 850degC, to convert uranium into triuranium octoxide. Uranium in the triuranium octoxide was determined gravimetrically. Ten grams of caoted fuel particles were completely decomposed by fusion with 50 g of potassium fluoride at 900degC for 3 hrs. Analytical result for uranium in the fuel compact by the proposed method was 21.04 ± 0.05 g (n = 3), and was in good agreement with that obtained by non-destructive γ-ray measurement method : 21.01 ± 0.07 g (n = 3). (author)

  6. Particle Image Velocimetry and Computational Fluid Dynamics Analysis of Fuel Cell Manifold

    DEFF Research Database (Denmark)

    Lebæk, Jesper; Blazniak Andreasen, Marcin; Andresen, Henrik Assenholm

    2010-01-01

    The inlet effect on the manifold flow in a fuel cell stack was investigated by means of numerical methods (computational fluid dynamics) and experimental methods (particle image velocimetry). At a simulated high current density situation the flow field was mapped on a 70 cell simulated cathode...

  7. International R and D project on development of coated particle fuel for innovative reactors

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    The paper presents an outline for an international collaborative project of coated particle fuel development for innovative reactors. Specific issues include identification of R and D needs and the Member State facilities for meeting the needs followed by development and demonstration of technology. (author)

  8. Interaction between UO2 kernel and pyrocarbon coating in irradiated and unirradiated HTR fuel particles

    International Nuclear Information System (INIS)

    Drago, A.; Klersy, R.; Simoni, O.; Schrader, K.H.

    1975-08-01

    Experimental observations on unidirectional UO 2 kernel migration in TRISO type coated particle fuels are reported. An analysis of the experimental results on the basis of data and models from the literature is reported. The stoichiometric composition of the kernel is considered the main parameter that, associated with a temperature gradient, controls the unidirectional kernel migration

  9. Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

    1995-03-01

    Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

  10. Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, Friederike [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulation requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference

  11. EXTENDCHAIN: a package of computer programs for calculating the buildup of heavy metals, fission products, and activation products in reactor fuel elements

    International Nuclear Information System (INIS)

    Robertson, M.W.

    1977-01-01

    Design of HTGR recycle and refabrication facilities requires a detailed knowledge of the concentrations of around 400 nuclides which are segregated into four different fuel particle types. The EXTENDCHAIN package of computer programs and the supporting input data files were created to provide an efficient method for calculating the 1600 different concentrations required. The EXTENDCHAIN code performs zero-dimensional nuclide burnup, decay, and activation calculations in nine energy groups for up to 108 nuclides per run. Preparation and handling of the input and output for the sixteen EXTENDCHAIN runs required to produce the desired data are the most time consuming tasks in the computation of the spent fuel element composition. The EXTENDCHAIN package of computer programs contains four codes to aid in the preparation and handling of these data. Most of the input data such as cross sections, decay constants, and the nuclide interconnection scheme will not change when calculating new cases. These data were developed for the life cycle of a typical HTGR and stored on archive tapes for future use. The fuel element composition for this typical HTGR life has been calculated and the results for an equilibrium recycle reload are presented. 12 figures, 7 tables

  12. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

  13. High-temperature process heat applications with an HTGR

    International Nuclear Information System (INIS)

    Quade, R.N.; Vrable, D.L.

    1980-04-01

    An 842-MW(t) HTGR-process heat (HTGR-PH) design and several synfuels and energy transport processes to which it could be coupled are described. As in other HTGR designs, the HTGR-PH has its entire primary coolant system contained in a prestressed concrete reactor vessel (PCRV) which provides the necessary biological shielding and pressure containment. The high-temperature nuclear thermal energy is transported to the externally located process plant by a secondary helium transport loop. With a capability to produce hot helium in the secondary loop at 800 0 C (1472 0 F) with current designs and 900 0 C (1652 0 F) with advanced designs, a large number of process heat applications are potentially available. Studies have been performed for coal liquefaction and gasification using nuclear heat

  14. HTGR safety research program. Progress report, April--June 1975

    International Nuclear Information System (INIS)

    Kirk, W.L.

    1975-09-01

    Progress in HTGR safety research is reported under the following headings: fission product technology; primary coolant impurities; structural investigation; safety instrumentation and control systems; phenomena modeling and systems analysis. (JWR)

  15. Status of the United States National HTGR program

    International Nuclear Information System (INIS)

    1981-01-01

    The HTGR continues to appear as an increasingly attractive option for application to US energy markets. To examine that potential, a program is being pursued to examine the various HTGR applications and to provide information to decision-makers in both the public and private sectors. To date, this effort has identified a substantial technical and economic potential for Steam Cycle/Cogeneration applications. Advanced HTGR systems are currently being evaluated to determine their appropriate role and timing. The encouraging results which have been obtained lead to heightened anticipation that a role for the HTGR will be found in the US energy market and that an initiative culminating in a lead project will be evolved in the forseeable future. The US Program can continue to benefit from international cooperative activities to develop the needed technologies. Expansion of these cooperative activities will be actively pursued

  16. Volume 2. Probabilistic analysis of HTGR application studies. Supporting data

    International Nuclear Information System (INIS)

    1980-09-01

    Volume II, Probabilistic Analysis of HTGR Application Studies - Supporting Data, gives the detail data, both deterministic and probabilistic, employed in the calculation presented in Volume I. The HTGR plants and the fossil plants considered in the study are listed. GCRA provided the technical experts from which the data were obtained by MAC personnel. The names of the technical experts (interviewee) and the analysts (interviewer) are given for the probabilistic data

  17. Technical review of process heat applications using the HTGR

    International Nuclear Information System (INIS)

    Brierley, G.

    1976-06-01

    The demand for process heat applications is surveyed. Those applications which can be served only by the high temperature gas-cooled reactor (HTGR) are identified and the status of process heat applications in Europe, USA, and Japan in December 1975 is discussed. Technical problems associated with the HTGR for process heat applications are outlined together with an appraisal of the safety considerations involved. (author)

  18. HTGR high temperature process heat design and cost status report

    International Nuclear Information System (INIS)

    1981-12-01

    This report describes the status of the studies conducted on the 850 0 C ROT indirect cycle and the 950 0 C ROT direct cycle through the end of Fiscal Year 1981. Volume I provides summaries of the design and optimization studies and the resulting capital and product costs, for the HTGR/thermochemical pipeline concept. Additionally, preliminary evaluations are presented for coupling of candidate process applications to the HTGR system

  19. Assessment of the licensing aspects of HTGR in Yugoslavia

    International Nuclear Information System (INIS)

    Varazdinec, Z.

    1990-01-01

    This paper deals not only with the licensing procedure in Yugoslavia, but also reflects the Utility/Owner approach to the assessment of the licensability of the HTGR during the site selection process and especially during bid evaluation process. Besides the description of the existing procedure which was implemented on licensing of LWR program, the assessment of some licensing aspects of HTGR has been presented to describe possible implementation on licensing procedure. (author)

  20. Assessment of the licensing aspects of HTGR in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Varazdinec, Z [Institut za Elektroprivredu-Zagreb, Zagreb (Yugoslavia)

    1990-07-01

    This paper deals not only with the licensing procedure in Yugoslavia, but also reflects the Utility/Owner approach to the assessment of the licensability of the HTGR during the site selection process and especially during bid evaluation process. Besides the description of the existing procedure which was implemented on licensing of LWR program, the assessment of some licensing aspects of HTGR has been presented to describe possible implementation on licensing procedure. (author)

  1. Burn-up calculations for a thorium HTR with one and with two types of fuel particle

    Energy Technology Data Exchange (ETDEWEB)

    Griggs, C. F.

    1975-06-15

    Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

  2. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-01-01

    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. Near-term development is focused on electricity generation. The top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. Quantitatively, this requires that the design meet the US Environmental Protection Agency's Protective Action Guides at the site boundary and hence preclude the need for sheltering or evacuation of the public. To meet these stringent safety requirements and at the same time provide a cost competitive design requires the innovative use of the basic high temperature gas-cooled reactor features of ceramic fuel, helium coolant, and a graphite moderator. The specific fuel composition and core size and configuration have been selected to the use the natural characteristics of these materials to develop significantly higher margins of safety. In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure of a simultaneous loss of forced cooling of the core. The challenges to control of heat generation consider not only the failure to insert the reactivity control systems but also the withdrawal of control rods. Finally, challenges to control of chemical attack of the ceramic-coated fuel are considered, including catastrophic failure of the steam generator, which allows water ingress, or failure of the pressure vessels, which allows air ingress. The plant's response to these extreme challenges is not dependent on operator action, and the events considered encompass conceivable operator errors

  3. Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts

    International Nuclear Information System (INIS)

    Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

    2014-01-01

    As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of "1"3"4Cs and "1"3"7Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during the AGR-1 irradiation test or post-irradiation safety testing at 1600– 1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the ORNL Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking. (author)

  4. Long discharge particle balance and fuel retention in Tore Supra

    International Nuclear Information System (INIS)

    Pegourie, B.; Brosset, C.; Delchambre, E.; Loarer, T.; Tsitrone, E.; Bucalossi, J.; Gunn, J.; Reichle, R.; Khodja, H.; Lafon, C.; Parent, P.

    2003-01-01

    In the new CIEL configuration of Tore-Supra, all the plasma facing components are actively cooled. The surface of the wall covered with carbon is about 15 m 2 (2-4 m 2 of which in close interaction with the plasma). This configuration allowed to maintain steady-state plasma conditions during up to 4 min 25 s. In these experiments, the required gas injection for maintaining the prescribed density remains constant during the whole discharge. The exhausted flux is also constant and equal to 40-50% of the injected flux. Therefore, 50-60% of the injected particles remain trapped in the vessel, the total retention being proportional to the plasma duration. Since the gas recovered between shots or by conditioning techniques is far from balancing the injected gas, huge quantities of deuterium remain indefinitely trapped in the wall, which appears as an infinite reservoir. This reservoir is believed to be dominated by co-deposited layers, as observed in several places of the vessel. The thickest deposits (up to 800 μm) are observed on the leading edge of the neutralizers of the exhaust system. They display a column-like shape (typical growth rate ∼ 20 nm/s) and have a graphite-like structure. Their deuterium content is modest. Conversely, when coming from regions protected from the direct plasma flux, the deposits show a smoother shape and their deuterium content is typically an order of magnitude larger. (authors)

  5. Volume 1. Probabilistic analysis of HTGR application studies. Technical discussion

    International Nuclear Information System (INIS)

    May, J.; Perry, L.

    1980-01-01

    The HTGR Program encompasses a number of decisions facing both industry and government which are being evaluated under the HTGR application studies being conducted by the GCRA. This report is in support of these application studies, specifically by developing comparative probabilistic energy costs of the alternative HTGR plant types under study at this time and of competitive PWR and coal-fired plants. Management decision analytic methodology was used as the basis for the development of the comparative probabilistic data. This study covers the probabilistic comparison of various HTGR plant types at a commercial development stage with comparative PWR and coal-fired plants. Subsequent studies are needed to address the sequencing of HTGR plants from the lead plant to the commercial plants and to integrate the R and D program into the plant construction sequence. The probabilistic results cover the comparison of the 15-year levelized energy costs for commercial plants, all with 1995 startup dates. For comparison with the HTGR plants, PWR and fossil-fired plants have been included in the probabilistic analysis, both as steam electric plants and as combined steam electric and process heat plants

  6. Effects of fuel components and combustion particle physicochemical properties on toxicological responses of lung cells.

    Science.gov (United States)

    Jaramillo, Isabel C; Sturrock, Anne; Ghiassi, Hossein; Woller, Diana J; Deering-Rice, Cassandra E; Lighty, JoAnn S; Paine, Robert; Reilly, Christopher; Kelly, Kerry E

    2018-03-21

    The physicochemical properties of combustion particles that promote lung toxicity are not fully understood, hindered by the fact that combustion particles vary based on the fuel and combustion conditions. Real-world combustion-particle properties also continually change as new fuels are implemented, engines age, and engine technologies evolve. This work used laboratory-generated particles produced under controlled combustion conditions in an effort to understand the relationship between different particle properties and the activation of established toxicological outcomes in human lung cells (H441 and THP-1). Particles were generated from controlled combustion of two simple biofuel/diesel surrogates (methyl decanoate and dodecane/biofuel-blended diesel (BD), and butanol and dodecane/alcohol-blended diesel (AD)) and compared to a widely studied reference diesel (RD) particle (NIST SRM2975/RD). BD, AD, and RD particles exhibited differences in size, surface area, extractable chemical mass, and the content of individual polycyclic aromatic hydrocarbons (PAHs). Some of these differences were directly associated with different effects on biological responses. BD particles had the greatest surface area, amount of extractable material, and oxidizing potential. These particles and extracts induced cytochrome P450 1A1 and 1B1 enzyme mRNA in lung cells. AD particles and extracts had the greatest total PAH content and also caused CYP1A1 and 1B1 mRNA induction. The RD extract contained the highest relative concentration of 2-ring PAHs and stimulated the greatest level of interleukin-8 (IL-8) and tumor necrosis factor-alpha (TNFα) cytokine secretion. Finally, AD and RD were more potent activators of TRPA1 than BD, and while neither the TRPA1 antagonist HC-030031 nor the antioxidant N-acetylcysteine (NAC) affected CYP1A1 or 1B1 mRNA induction, both inhibitors reduced IL-8 secretion and mRNA induction. These results highlight that differences in fuel and combustion conditions

  7. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  8. Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

    Science.gov (United States)

    Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

    2017-01-01

    This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

  9. Error Analysis of Ceramographic Sample Preparation for Coating Thickness Measurement of Coated Fuel Particles

    International Nuclear Information System (INIS)

    Liu Xiaoxue; Li Ziqiang; Zhao Hongsheng; Zhang Kaihong; Tang Chunhe

    2014-01-01

    The thicknesses of four coatings of HTR coated fuel particle are very important parameters. It is indispensable to control the thickness of four coatings of coated fuel particles for the safety of HTR. A measurement method, ceramographic sample-microanalysis method, to analyze the thickness of coatings was developed. During the process of ceramographic sample-microanalysis, there are two main errors, including ceramographic sample preparation error and thickness measurement error. With the development of microscopic techniques, thickness measurement error can be easily controlled to meet the design requirements. While, due to the coated particles are spherical particles of different diameters ranged from 850 to 1000μm, the sample preparation process will introduce an error. And this error is different from one sample to another. It’s also different from one particle to another in the same sample. In this article, the error of the ceramographic sample preparation was calculated and analyzed. Results show that the error introduced by sample preparation is minor. The minor error of sample preparation guarantees the high accuracy of the mentioned method, which indicates this method is a proper method to measure the thickness of four coatings of coated particles. (author)

  10. Fuel Design for Particle-Bed Reactors for Thermal Propulsion Applications

    Science.gov (United States)

    Husser, Dewayne L.; Evans, Robert S.; Jensen, Russell R.; Kerr, John M.

    1994-07-01

    The design of particle bed reactor (PBR) fuels is an iterative process involving close coordination of design and manufacturing operations. The process starts with the generation of an initial particle design, based on a knowledge of the system requirements and interfaces (such as, fissile loading requirements, coolant type, exit gas temperatures, operation time, number of cycles, contacting materials, etc.). The designer must consider materials property data, heat-transfer and thermal-hydraulic characteristics of the particle and particle bed, and available (or anticipated) manufacturing technology. The design process also uses parametric studies to identify the influences of composition, size, and coating thickness on fuel performance. This resulting design is then used to provide a target manufacturing specification against which initial manufacturing development can be assessed and which provides the framework for manufacturing and testing derived feedback that can be incorporated into the subsequent particle design modifications. In this paper, an example of this design process for a hypothetical particle using a (U,Zr)C kernel and a NbC outer coating designed for a thermal propulsion application is given.

  11. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

    Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

    1990-01-01

    Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

  12. Screening of synfuel processes for HTGR application

    International Nuclear Information System (INIS)

    1981-02-01

    The aim of this study is to select for further study, the several synfuel processes which are the most attractive for application of HTGR heat and energy. In pursuing this objective, the Working Group identified 34 candidate synfuel processes, cut the number of processes to 16 in an initial screening, established 11 prime criteria with weighting factors for use in screening the remaining processes, developed a screening methodology and assumptions, collected process energy requirement information, and performed a comparative rating of the processes. As a result of this, three oil shale retorting processes, two coal liquefaction processes and one coal gasification process were selected as those of most interest for further study at this time

  13. The acoustic environment in large HTGR's

    International Nuclear Information System (INIS)

    Burton, T.E.

    1979-01-01

    Well-known techniques for estimating acoustic vibration of structures have been applied to a General Atomic high-temperature gas-cooled reactor (HTGR) design. It is shown that one must evaluate internal loss factors for both fluid and structure modes, as well as radiation loss factors, to avoid large errors in estimated structural response. At any frequency above 1350 rad/s there are generally at least 20 acoustic modes contributing to acoustic pressure, so statistical energy analysis may be employed. But because the gas circuit consists mainly of high-aspect-ratio cavities, reverberant fields are nowhere isotropic below 7500 rad/s, and in some regions are not isotropic below 60 000 rad/s. In comparison with isotropic reverberant fields, these anistropic fields enhance the radiation efficiencies of some structural modes at low frequencies, but have surprisingly little effect at most frequencies. The efficiency of a dipole sound source depends upon its orientation. (Auth.)

  14. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Wang, Lumin; Was, Gary

    2010-01-01

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  15. 1170-MW(t) HTGR-PS/C plant application study report: tar sands oil recovery application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to tar sands oil recovery and upgrading. The raw product recovered from the sands is a heavy, sour bitumen; upgrading, which involves coking and hydrodesulfurization, produces a synthetic crude (refinable by current technology) and petroleum coke. Steam and electric power are required for the recovery and upgrading process. Proposed and commercial plants would purchase electric power from local utilities and obtain from boilers fired with coal and with by-product fuels produced by the upgrading. This study shows that an HTGR-PS/C represents a more economical source of steam and electric power

  16. Particle and carbon dioxide emissions from passenger vehicles operating on unleaded petrol and LPG fuel

    International Nuclear Information System (INIS)

    Ristovski, Z.D.; Jayaratne, E.R.; Morawska, L.; Ayoko, G.A.; Lim, M.

    2005-01-01

    A comprehensive study of the particle and carbon dioxide emissions from a fleet of six dedicated liquefied petroleum gas (LPG) powered and five unleaded petrol (ULP) powered new Ford Falcon Forte passenger vehicles was carried out on a chassis dynamometer at four different vehicle speeds-0 (idle), 40, 60, 80 and 100 km h -1 . Emission factors and their relative values between the two fuel types together with a statistical significance for any difference were estimated for each parameter. In general, LPG was found to be a 'cleaner' fuel, although in most cases, the differences were not statistically significant owing to the large variations between emissions from different vehicles. The particle number emission factors ranged from 10 11 to 10 13 km -1 and was over 70% less with LPG compared to ULP. Corresponding differences in particle mass emission factor between the two fuels were small and ranged from the order of 10 μg km -1 at 40 to about 1000 μg km -1 at 100 km h -1 . The count median particle diameter (CMD) ranged from 20 to 35 nm and was larger with LPG than with ULP in all modes except the idle mode. Carbon dioxide emission factors ranged from about 300 to 400 g km -1 at 40 km h -1 , falling with increasing speed to about 200 g km -1 at 100 km h -1 . At all speeds, the values were 10% to 18% greater with ULP than with LPG

  17. The influence of design and fuel parameters on the particle emissions from wood pellets combustion. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wiinikka, Henrik; Gebart, Rikard [Energy Technology Centre, Piteaa (Sweden)

    2005-02-01

    Combustion of solid biomass under fixed bed conditions is a common technique to generate heat and power in both small and large scale grate furnaces (domestic boilers, stoves, district heating plants). Unfortunately, combustion of biomass will generate particle emissions containing both large fly ash particles and fine particles that consist of fly ash and soot. The large fly ash particles have been produced from fusion of non-volatile ash-forming species in burning char particle. The inorganic fine particles have been produced from nucleation of volatilised ash elements (K, Na, S, Cl and Zn). If the combustion is incomplete, soot particles are also produced from secondary reaction of tar. The particles in the fine fraction grows by coagulation and coalescence to a particle diameter around 0.1 pm. Since the smallest particles are very hard to collect in ordinary cleaning devices they contribute to the ambient air pollution. Furthermore, fine airborne particles have been correlated to adverse effects on the human health. It is therefore essential to minimize particle formation from the combustion process and thereby reduce the emissions of particulates to the ambient air. The aim with this project is to study particle emissions from small scale combustion of wood pellets and to investigate the impact of different operating, construction and fuel parameters on the amount and characteristic of the combustion generated particles. To address these issues, experiments were carried out in a 10 kW updraft fired wood pellets reactor that has been custom designed for systematic investigations of particle emissions. In the flue gas stack, particle emissions were sampled on a filter. The particle mass and number size distributions were analysed by a low pressure cascade impactor and a SMPS (Scanning Electron Mobility Particle Sizer). The results showed that the temperature and the flow pattern in the combustion zone affect the particle emissions. Increasing combustion

  18. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    International Nuclear Information System (INIS)

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

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600 0 C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth

  19. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  20. Morphology of single inhalable particle inside public transit biodiesel fueled bus.

    Science.gov (United States)

    Shandilya, Kaushik K; Kumar, Ashok

    2010-01-01

    In an urban-transit bus, fueled by biodiesel in Toledo, Ohio, single inhalable particle samples in October 2008 were collected and detected by scanning electron microscopy and energy dispersive X-ray spectrometry (SEM/EDS). Particle size analysis found bimodal distribution at 0.2 and 0.5 microm. The particle morphology was characterized by 14 different shape clusters: square, pentagon, hexagon, heptagon, octagon, nonagon, decagon, agglomerate, sphere, triangle, oblong, strip, line or stick, and unknown, by quantitative order. The square particles were common in the samples. Round and triangle particles are more, and pentagon, hexagon, heptagon, octagon, nonagon, decagon, strip, line or sticks are less. Agglomerate particles were found in abundance. The surface of most particles was coarse with a fractal edge that can provide a suitable chemical reaction bed in the polluted atmospheric environment. The three sorts of surface patterns of squares were smooth, semi-smooth, and coarse. The three sorts of square surface patterns represented the morphological characteristics of single inhalable particles in the air inside the bus in Toledo. The size and shape distribution results were compared to those obtained for a bus using ultra low sulfur diesel.

  1. Effects of graphite surface roughness on bypass flow computations for an HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Yu-Hsin, E-mail: touushin@gmail.com [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Johnson, Richard W., E-mail: Rich.Johnson@inl.gov [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Sato, Hiroyuki, E-mail: sato.hiroyuki09@jaea.go.jp [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer CFD calculations are made of bypass flow between graphite blocks in HTGR. Black-Right-Pointing-Pointer Several turbulence models are employed to compare to friction and heat transfer correlations. Black-Right-Pointing-Pointer Parameters varied include bypass gap width and surface roughness. Black-Right-Pointing-Pointer Surface roughness causes increases in max fuel and coolant temperatures. Black-Right-Pointing-Pointer Surface roughness does not cause increase in outlet coolant temperature variation. - Abstract: Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow; it has been suggested that it may be as much as 20% of the total helium coolant flow [INL Report 2007, INL/EXT-07-13289]. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors for three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U.S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for steady flow in a

  2. New developments in image-based characterization of coated particle nuclear fuel

    Science.gov (United States)

    Price, Jeffery R.; Aykac, Deniz; Hunn, John D.; Kercher, Andrew K.; Morris, Robert N.

    2006-02-01

    We describe in this paper new developments in the characterization of coated particle nuclear fuel using optical microscopy and digital imaging. As in our previous work, we acquire optical imagery of the fuel pellets in two distinct manners that we refer to as shadow imaging and cross-sectional imaging. In shadow imaging, particles are collected in a single layer on an optically transparent dish and imaged using collimated back-lighting to measure outer surface characteristics only. In cross-sectional imaging, particles are mounted in acrylic epoxy and polished to near-center to reveal the inner coating layers for measurement. For shadow imaging, we describe a curvaturebased metric that is computed from the particle boundary points in the FFT domain using a low-frequency parametric representation. We also describe how missing boundary points are approximated using band-limited interpolation so that the FFT can be applied. For cross-section imaging, we describe a new Bayesian-motivated segmentation scheme as well as a new technique to correct layer measurements for the fact that we cannot observe the true mid-plane of the approximately spherical particles.

  3. Comparison of stochastic models in Monte Carlo simulation of coated particle fuels

    International Nuclear Information System (INIS)

    Yu Hui; Nam Zin Cho

    2013-01-01

    There is growing interest worldwide in very high temperature gas cooled reactors as candidates for next generation reactor systems. For design and analysis of such reactors with double heterogeneity introduced by the coated particle fuels that are randomly distributed in graphite pebbles, stochastic transport models are becoming essential. Several models were reported in the literature, such as coarse lattice models, fine lattice stochastic (FLS) models, random sequential addition (RSA) models, metropolis models. The principles and performance of these stochastic models are described and compared in this paper. Compared with the usual fixed lattice methods, sub-FLS modeling allows more realistic stochastic distribution of fuel particles and thus results in more accurate criticality calculation. Compared with the basic RSA method, sub-FLS modeling requires simpler and more efficient overlapping checking procedure. (authors)

  4. In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

    Science.gov (United States)

    Terrani, K. A.; Silva, C. M.; Kiggans, J. O.; Cai, Z.; Shin, D.; Snead, L. L.

    2013-06-01

    The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was the least substantial, while the PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.

  5. Diffusivities of Ag, Cs, Sr, and Kr in TRISO fuel particles and graphite

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Tri-structural isotropic (TRISO) coated particles have been developed and studied since the late 1950s when the concept of coated particles was invented by Roy Huddle of the United Kingdom Atomic Energy Authority. Several decades of work by half a dozen countries on fission product transport in TRISO fuel through numerous irradiation and heating experiments have led to several recommendations of transport data and to the adoption of various sets of diffusion coefficients. In 1997, the International Atomic Energy Agency (IAEA) gathered all these historical results and issued a technical document (TECDOC-978 [IAEA]) that summarizes these sets of recommended diffusion coefficients. Table 1 shows the reference literature articles for the diffusivities that have historically been recommended by the American and German TRISO fuel development programs and that are summarized in the IAEA report (see section 7 for full references of these articles).

  6. Thermo-mechanical behaviour modelling of particle fuels using a multi-scale approach

    International Nuclear Information System (INIS)

    Blanc, V.

    2009-12-01

    Particle fuels are made of a few thousand spheres, one millimeter diameter large, compound of uranium oxide coated by confinement layers which are embedded in a graphite matrix to form the fuel element. The aim of this study is to develop a new simulation tool for thermo-mechanical behaviour of those fuels under radiations which is able to predict finely local loadings on the particles. We choose to use the square finite element method, in which two different discretization scales are used: a macroscopic homogeneous structure whose properties in each integration point are computed on a second heterogeneous microstructure, the Representative Volume Element (RVE). First part of this works is concerned by the definition of this RVE. A morphological indicator based in the minimal distance between spheres centers permit to select random sets of microstructures. The elastic macroscopic response of RVE, computed by finite element has been compared to an analytical model. Thermal and mechanical representativeness indicators of local loadings has been built from the particle failure modes. A statistical study of those criteria on a hundred of RVE showed the significance of choose a representative microstructure. In this perspective, a empirical model binding morphological indicator to mechanical indicator has been developed. Second part of the work deals with the two transition scale method which are based on the periodic homogenization. Considering a linear thermal problem with heat source in permanent condition, one showed that the heterogeneity of the heat source involve to use a second order method to localized finely the thermal field. The mechanical non-linear problem has been treats by using the iterative Cast3M algorithm, substituting to integration of the behavior law a finite element computation on the RVE. This algorithm has been validated, and coupled with thermal resolution in order to compute a radiation loading. A computation on a complete fuel element

  7. Anisotropy variation of crystallographic orientation in pyrocarbon coatings of fuel particles by annealing and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Koizlik, K.

    1973-04-15

    This document is a translation of those parts of the German report Jul-868-RW concerned with changes in anisotropy as determined using an optical technique on pyrocarbon coatings on fuel particles resulting from annealing and neutron irradiations. Two lists of contents are included, one is for the present document and the other is the full contents of the original report and is included for the generl interest of users.

  8. CHAP: a composite nuclear plant simulation program applied to the 3000 MW(t) HTGR

    International Nuclear Information System (INIS)

    Secker, P.A.; Bailey, P.G.; Gilbert, J.S.; Willcutt, G.J.E. Jr.; Vigil, J.C.

    1977-01-01

    The Composite HTGR Analysis Program (CHAP) is a general systems analysis program which has been developed at LASL. The program is being used for simulating large HTGR nuclear power plant operation and accident transients. The general features and analytical methods of the CHAP program are discussed. Features of the large HTGR model and results of model transients are also presented

  9. Present status of HTGR projects and their heat applications in Russia

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Glushkov, E.S.; Kukharkin, N.E.; Ponomarev-Stepnoi, N.N.

    1996-01-01

    This paper describes the main technical decision and parameters of the HTGR of different power and considers a few schemes of HTGR plants with a gas turbine cycle. Also, the future prospects on heat utilization of HTGR in Russia is presented. (J.P.N.)

  10. Solubility of hot fuel particles from Chernobyl--influencing parameters for individual radiation dose calculations.

    Science.gov (United States)

    Garger, Evgenii K; Meisenberg, Oliver; Odintsov, Oleksiy; Shynkarenko, Viktor; Tschiersch, Jochen

    2013-10-15

    Nuclear fuel particles of Chernobyl origin are carriers of increased radioactivity (hot particles) and are still present in the atmosphere of the Chernobyl exclusion zone. Workers in the zone may inhale these particles, which makes assessment necessary. The residence time in the lungs and the transfer in the blood of the inhaled radionuclides are crucial for inhalation dose assessment. Therefore, the dissolution of several kinds of nuclear fuel particles from air filters sampled in the Chernobyl exclusion zone was studied. For this purpose filter fragments with hot particles were submersed in simulated lung fluids (SLFs). The activities of the radionuclides (137)Cs, (90)Sr, (239+240)Pu and (241)Am were measured in the SLF and in the residuum of the fragments by radiometric methods after chemical treatment. Soluble fractions as well as dissolution rates of the nuclides were determined. The influence of the genesis of the hot particles, represented by the (137)Cs/(239+240)Pu ratio, on the availability of (137)Cs was demonstrated, whereas the dissolution of (90)Sr, (239+240)Pu and (241)Am proved to be independent of genesis. No difference in the dissolution of (137)Cs and (239+240)Pu was observed for the two applied types of SLF. Increased solubility was found for smaller hot particles. A two-component exponential model was used to describe the dissolution of the nuclides as a function of time. The results were applied for determining individual inhalation dose coefficients for the workers at the Chernobyl construction site. Greater dose coefficients for the respiratory tract and smaller coefficients for the other organs were calculated (compared to ICRP default values). The effective doses were in general lower for the considered radionuclides, for (241)Am even by one order of magnitude. © 2013 Elsevier B.V. All rights reserved.

  11. Approach on a global HTGR R and D network

    International Nuclear Information System (INIS)

    Lensa, W. von

    1997-01-01

    The present situation of nuclear power in general and of the innovative nuclear reactor systems in particular requires more comprehensive, coordinated R and D efforts on a broad international level to respond to today's requirements with respect to public and economic acceptance as well as to globalization trends and global environmental problems. HTGR technology development has already reached a high degree of maturity that will be complemented by the operation of the two new test reactors in Japan and China, representing technological milestones for the demonstration of HTGR safety characteristics and Nuclear Process Heat generation capabilities. It is proposed by the IAEA 'International Working Group on Gas-Cooled Reactors' to establish a 'Global HTGR R and D Network' on basic HTGR technology for the stable, long-term advancement of the specific HTGR features and as a basis for the future market introduction of this innovative reactor system. The background and the motivation for this approach are illustrated, as well as first proposals on the main objectives, the structure and the further procedures for the implementation of such a multinational working sharing R and D network. Modern telecooperation methods are foreseen as an interactive tool for effective communication and collaboration on a global scale. (author)

  12. New developments in JET neutron, alpha particle and fuel mixture diagnostics with potential relevance to ITER

    International Nuclear Information System (INIS)

    Murari, A.; Bertalot, L.; Angelone, M.; Pillon, M.; Ericsson, G.; Conroy, S.; Kaellne, J.; Kiptily, V.; Popovichev, S.; Adams, J.M.; Stork, D.; Afanasyiev, V.; Mironov, M.; Bonheure, G.

    2005-01-01

    Some recent JET campaigns, with the introduction of trace amount (n T /n D 4 He, provided unique opportunities to test new diagnostic approaches and technologies for the detection of neutrons, alpha particles and fuel mixture. With regard to neutron detection, the recent activity covered all the most essential aspects: calibration and cross validation of the diagnostics, measurement of the spatial distribution of the neutrons, particle transport and finally neutron spectrometry. The first tests of some new neutron detection technologies were also undertaken successfully during the TTE campaign. To improve JET diagnostic capability in the field of alpha particles, a strong development program was devoted to the measurement of their slowing down and imaging with gamma ray spectroscopy. A new approach for the fusion community to measure the fast ion losses, based on the activation technique, was also successfully attempted for the first time on JET. A careful assessment of the NPA potential to determine the fuel mixture and the particle transport coefficients is under way. (author)

  13. Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

    International Nuclear Information System (INIS)

    Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J.; Meyer, Willem C.H.M.; Hyatt, Neil C.

    2013-01-01

    Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H 2 /N 2 atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required

  14. Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom); Meyer, Willem C.H.M. [Necsa, South African Nuclear Energy Corporation, PO Box 582, Pretoria, Gauteng (South Africa); Hyatt, Neil C., E-mail: n.c.hyatt@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom)

    2013-05-15

    Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H{sub 2}/N{sub 2} atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required.

  15. Mass spectrometric determination of gases in individual coated HTR fuel particles. I

    International Nuclear Information System (INIS)

    Strigl, A.; Bildstein, H.

    1977-01-01

    A method is described which allows the simultaneous determination of fission and reaction gases in individual coated particles at temperatures up to 2 000 0 C. The particles are heated under high-vacuum in a micro resistance-furnace up to the desired temperature. After preselected times the particles are crushed by action of a pneumatic cylinder. The gases liberated are fed into a quadrupole analyzer where they are analyzed in a dynamic mode. A peak selector allows the simultaneous measurement of up to four gases. The method is used routinely for the determination of fission gases (Kr and Xe) and of carbon monoxide which is formed as a reaction gas from oxide fuel. Precision and accuracy are in the order of a few percent. Detection limits for routine measurements are about 10 -7 cm 3 (STP) for Kr and Xe and 2x10 -5 cm 3 (STP) for CO but can be lowered by special techniques. (Auth.)

  16. Fission product Pd-SiC interaction in irradiated coated particle fuels

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1980-04-01

    Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO 2 to UC 2 , including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% 235 U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary

  17. 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

  18. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  19. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  20. Subharmonic excitation in an HTGR core

    International Nuclear Information System (INIS)

    Bezler, P.; Curreri, J.R.

    1977-01-01

    The occurrence of subharmonic resonance in a series of blocks with clearance between blocks and with springs on the outer most ends is the subject of this paper. This represents an HTGR core response to an earthquake input. An analytical model of the cross section of this type of core is a series of blocks arranged horizontally between outer walls. Each block represents many graphite hexagonal core elements acting in unison as a single mass. The blocks are of unequal size to model the true mass distribution through the core. Core element elasticity and damping characteristics are modeled with linear spring and viscous damping units affixed to each block. The walls and base represent the core barell or core element containment structure. For forced response calculations, these boundaries are given prescribed motions. The clearance between each block could be the same or different with the total clearance duplicating that of the entire core. Spring packs installed between the first and last block and the boundaries model the boundary elasticity. The system non-linearity is due to the severe discontinuity in the interblock elastic forces when adjacent blocks collide. A computer program using a numerical integration scheme was developed to solve for the response of the system to arbitrary inputs