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HTGR Measurements and Instrumentation Systems

Description: 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.
Date: May 2012
Creator: Ball, Sydney J.; Holcomb, David Eugene & Cetiner, Mustafa Sacit
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Improved size uniformity of sol-gel spheres by imposing a vibration on the sol in dispersion nozzles

Description: A major part of the Th-- /sup 233/U fuel cycle program at ORNL has been concerned with the development of sol-gel processes to prepare ThO/sub 2/ and ThO/ sub 2/--/sup 233/UO/sub 2/ spheres. The formation of sol drops having a uniform and controlled diameter is important to any sol-gel process for preparing oxide kernels for High Temperature Gas-Cooled Reactor fuels. A recently developed technique that incorporates use of a sol disperser with vibration has met the dispersion requirements for ORNL sol-gel processes better than any of the previous techniques employing dispersers alone. With this new technique, the breakup of sol streams from orifices or capillaries is made more uniform and regular by imposing a vibration, at the natural frequency of drop formation, on the sol at the entrance to the orifices or capillaries. This techn-que has been applied to two-fluid nozzles to form 1,000 to 88,000 sol drops per minute. Batches consisting of 1 to 13 kg of fired ThO/sub 2/ spheres 370 to 500 mu in diameter had average diameters within 1% of the predicted values and standard deviations of 2.5 to 5.0 mu . Yields after both size and shape separation were greater than 95% and were usually greater than 98% in a 30- mu range. The same technique was applied to shear nozzles at rates up to 192,000 drops per minute. Nozzles of this type use multiple orifices and are more dependable for larger capacities than multiple two-fluid nozzles. (auth)
Date: May 1, 1973
Creator: Haas, P. A. & Lackey, W. J.
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Partner: UNT Libraries Government Documents Department

Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

Description: This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.
Date: May 1, 1980
Creator: Barsell, A.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

HTGR Gas Turbine Program. Semiannual progress report for the period ending September 30, 1979

Description: Information on the HTGR-GT program is presented concerning systems design methods; systems dynamics methods; alternate design; miscellaneous controls and auxiliary systems; structural mechanics; shielding analysis; licensing; safety; availability; reactor turbine system integration with plant; PCRV liners, penetrations, and closures; PCRV structures; thermal barrier; reactor internals; turbomachinery; turbomachine remote maintenance; control valve; heat exchangers; plant protection system; and plant control system.
Date: May 1, 1980
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Partner: UNT Libraries Government Documents Department

Comparison of calculated results from two analytical models with measured data from a heat-exchanger flow test

Description: Predicted results from both a network flow model and a turbulent flow model were compared with measured results from an air flow test on a half-scale model of the auxiliary heat exchanger for a high-temperature gas-cooled reactor. Measurements of both velocity and pressure were made within the heat exchanger shell side flow field. These measurements were compared with calculated results from both a network flow model and a turbulent flow model. Both analytical models predicted early identical results which, except for some minor anomalies, compared favorably with the measured data.
Date: May 1, 1983
Creator: Carosella, D.P. & Pavlics, P.N.
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Partner: UNT Libraries Government Documents Department

350 MW(t) MHTGR preassembly and modularization

Description: The Modular High Temperature Gas Cooled Reactor (MHTGR) provides a safe and economical nuclear power option for the world's electrical generation needs by the turn of the century. The proposed MHTGR plant is composed of four 350 MW(t) prismatic core reactor modules, coupled to a 2(2 {times} 1) turbine generator producing a net plant electrical output of 538 MW(e). Each of the four reactor module is located in a below-ground level concrete silo, and consists of a reactor vessel and a steam generator vessel interconnected by a cross duct vessel. The modules, along with the service buildings, are contained within a Nuclear Island (NI). The turbine generators and power generation facilities are in the non-nuclear Energy Conversion Area (ECA). The MHTGR design reduces cost and improves schedule by maximizing shop fabrication, minimizing field fit up of the Reactor Internals components and modularizing the NI ECA facilities. 3 refs., 6 figs., 2 tabs.
Date: May 1, 1991
Creator: Venkatesh, M.C. (General Atomics, San Diego, CA (USA)); Jones, G. (Gas-Cooled Reactor Associates, San Diego, CA (USA)); Dilling, D.A. (Bechtel International Corp., San Francisco, CA (USA)) & Parker, W.J. (Stone and Webster Engineering Corp., Boston, MA (USA))
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Partner: UNT Libraries Government Documents Department

Design of the HTGR for process heat applications

Description: This paper discusses a design study of an advanced 842-MW(t) HTGR with a reactor outlet temperature of 850/sup 0/C (1562/sup 0/F), coupled with a chemical process whose product is hydrogen (or a mixture of hydrogen and carbon monoxide) generated by steam reforming of a light hydrocarbon mixture. This paper discusses the plant layout and design for the major components of the primary and secondary heat transfer systems. Typical parametric system study results illustrate the capability of a computer code developed to model the plant performance and economics.
Date: May 1, 1980
Creator: Vrable, D.L. & Quade, R.N.
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Nuclear heat source component design considerations for HTGR process heat reactor plant concept

Description: 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.
Date: May 1, 1982
Creator: McDonald, C.F.; Kapich, D.; King, J.H. & Venkatesh, M.C.
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1170-MW(t) HTGR-PS/C plant application-study report: alumina-plant application

Description: This report considers the HTGR-PS/C application to producing alumina from bauxite. For the size alumina plant considered, the 1170-MW(t) HTGR-PS/C supplies 100% of the process steam and electrical power requirements and produces surplus electrical power and/or process steam, which can be used for other process users or electrical power production. Presently, the bauxite ore is reduced to alumina in plants geographically separated from the electrolysis plant. The electrolysis plants are located near economical electric power sources. However, with the integration of an 1170-MW(t) HTGR-PS/C unit in a commercial alumina plant, the excess electric power available (approx. 233 MW(e)) could be used for alumina electrolysis.
Date: May 1, 1981
Creator: Rao, R.; McMain, A.T. Jr. & Stanley, J.D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Gaseous and metallic fission product release characteristics of a modular pebble bed HTGR during loss of core cooling accidents

Description: A quantitative safety criteria for the high-temperature gas-cooled reactor (HTGR) is to limit the radiological consequences for a wide spectrum of accidents to a level not requiring public sheltering. This leads to reliance on passive safety characteristics for improbable loss of core cooling accidents. Models have been developed to predict the transport of metallic and gaseous fission products (FPs) through the multilayered fuel particle coatings and the graphite matrix of the core under accident conditions. Using these models, FP transport and releases were calculated for a loss of core convective cooling accident in a 250-MW(t) 3.8-W/cc pebble bed HTGR. Fission-product transport through the particle kernel and coatings, the graphite pebbles/reflectors, the reactor vessel, and the confinement were assessed. The results of this study show that the most effective barrier to fission products is the coated fuel particle. The reactor vessel and the confinement provide additional attenuation for the small amount released from the core. The small release to the environment occurs over a period of days and is so low that the safety criterion of 5 rem thyroid dose (to avoid offsite sheltering) is satisfied with a margin of more than an order of magnitude. 6 figs.
Date: May 1, 1985
Creator: Inamati, S.B.; Richards, M.B. & Hoot, C.G.
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Partner: UNT Libraries Government Documents Department

HTGR Fuel performance basis

Description: 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/sup 0/C, and complete fuel failure occurs at 2660/sup 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.
Date: May 1, 1982
Creator: Shamasundar, B.I.; Stansfield, O.M. & Jensen, D.D.
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Partner: UNT Libraries Government Documents Department

HTGR analytical methods and design verification

Description: 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.
Date: May 1, 1982
Creator: Neylan, A.J. & Northup, T.E.
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HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

Description: 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.
Date: May 1, 1979
Creator: Strand, J.B.; Fields, D.E. & Kergis, C.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

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

Description: 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/sub 2/) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC/sub 2/ and UO/sub 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.
Date: May 1, 1983
Creator: Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W. & Stansfield, O.M.
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Partner: UNT Libraries Government Documents Department

Analysis and evaluation of recent operational experience from the Fort St. Vrain HTGR

Description: The Fort St. Vrain operating experience to be discussed here includes notable safety-related events which have occurred since late 1981 when ORNL was first contracted to provide technical assistance to AEOD. Earlier Fort St. Vrain operating experience through the time of successful full-power testing in November 1981 has been summarized by the licensee and the reactor vendor, GA Technologies, Inc. (GA), in papers presented at several different forums during 1982. In addition, extensive and very useful detailed evaluations of preoperational and startup testing and of the rise-to-power operating experience through completion of the first refueling outage in August 1979 have been compiled into a series of reports under the sponsorship of the Electric Power Research Institute (EPRI). Finally, the US Department of Energy's Fort St. Vrain Improvement Plan provides a summary of the major operational limits which have affected the plant since start-up. The events discussed here are categorized based on the major systems affected, namely, (1) primary system and reactor vessel, (2) electrical systems, and (3) the reactor building. In all cases to be discussed, the lessons to be learned are vigilance and prevention. These lessons translate into the need for the recognition and control of unexpected situations and of their potential for branching effects. At Fort St. Vrain, these lessons are found in the effects of moisture ingress, in the challenges experienced to the supply of essential electrical power, and in controlling the environment of the reactor building. 13 refs.
Date: May 1, 1985
Creator: Moses, D.L. & Lanning, W.D.
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Consistent linearization method for finite-element analysis of viscoelastic materials

Description: A method of formulating material models for viscoelastic analysis using the finite-element method is presented. The method, named consistent linearization, includes the influence of creep in the material stiffness in a theoretically ideal manner. This method has been applied to the linear viscoelastic analysis of graphite subject to irradiation. Previously, using the initial strain method, short time steps had been required to avoid a numerical instability associated with the rapid transient creep. Using the consistent linearization method a factor of 15 reduction in computer time was achieved for the same accuracy.
Date: May 1, 1983
Creator: Smith, P.D. & Pelessone, D.
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Partner: UNT Libraries Government Documents Department

USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10

Description: This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.
Date: May 1, 1986
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Fission product plateout/liftoff/washoff test plan. Revision 1

Description: A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.
Date: May 1, 1988
Creator: Acharya, R. & Hanson, D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Comparison of US/FRG accident condition fuel failure and release models

Description: Although there are many differences between the High-Temperature Gas Cooled Reactor (HTGR) concepts being developed in the US and the High Temperature Reactor (HTR) concepts in the Federal Republic of Germany (FRG), the coated fuel particles are very similar. Significant benefits are achievable through cooperative research and exchange of information and data on the fuel performance and radionuclide retention in the coated fuel particles. This draft report describes cooperative work on HTGR safety research as agreed to in the "USA/FRG Umbrella Agreement for Cooperation in GCR Development: Safety Research Subprogram Plan," specifically, this work was conducted under Project Work Statement (PWS) S-6 titled "Fission Product Retention in Fuel," 9 refs., 12 figs., 4 tabs.
Date: May 29, 1989
Creator: Bolin, J. & Dunn, T.
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Partner: UNT Libraries Government Documents Department

Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

Description: Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.
Date: May 1, 1990
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department