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

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.

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

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

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

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

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.

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)

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

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

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

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.

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

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)

Nuclear closed-cycle gas turbine (HTGR-GT): dry cooled commercial power plant studies

International Nuclear Information System (INIS)

McDonald, C.F.; Boland, C.R.

1979-11-01

Combining the modern and proven power conversion system of the closed-cycle gas turbine (CCGT) with an advanced high-temperature gas-cooled reactor (HTGR) results in a power plant well suited to projected utility needs into the 21st century. The gas turbine HTGR (HTGR-GT) power plant benefits are consistent with national energy goals, and the high power conversion efficiency potential satisfies increasingly important resource conservation demands. Established technology bases for the HTGR-GT are outlined, together with the extensive design and development program necessary to commercialize the nuclear CCGT plant for utility service in the 1990s. This paper outlines the most recent design studies by General Atomic for a dry-cooled commercial plant of 800 to 1200 MW(e) power, based on both non-intercooled and intercooled cycles, and discusses various primary system aspects. Details are given of the reactor turbine system (RTS) and on integrating the major power conversion components in the prestressed concrete reactor vessel

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

Fault diagnosis of generation IV nuclear HTGR components – Part II: The area error enthalpy–entropy graph approach

International Nuclear Information System (INIS)

Rand, C.P. du; Schoor, G. van

2012-01-01

Highlights: ► Different uncorrelated fault signatures are derived for HTGR component faults. ► A multiple classifier ensemble increases confidence in classification accuracy. ► Detailed simulation model of system is not required for fault diagnosis. - Abstract: The second paper in a two part series presents the area error method for generation of representative enthalpy–entropy (h–s) fault signatures to classify malfunctions in generation IV nuclear high temperature gas-cooled reactor (HTGR) components. The second classifier is devised to ultimately address the fault diagnosis (FD) problem via the proposed methods in a multiple classifier (MC) ensemble. FD is realized by way of different input feature sets to the classification algorithm based on the area and trajectory of the residual shift between the fault-free and the actual operating h–s graph models. The application of the proposed technique is specifically demonstrated for 24 single fault transients considered in the main power system (MPS) of the Pebble Bed Modular Reactor (PBMR). The results show that the area error technique produces different fault signatures with low correlation for all the examined component faults. A brief evaluation of the two fault signature generation techniques is presented and the performance of the area error method is documented using the fault classification index (FCI) presented in Part I of the series. The final part of this work reports the application of the proposed approach for classification of an emulated fault transient in data from the prototype Pebble Bed Micro Model (PBMM) plant. Reference data values are calculated for the plant via a thermo-hydraulic simulation model of the MPS. The results show that the correspondence between the fault signatures, generated via experimental plant data and simulated reference values, are generally good. The work presented in the two part series, related to the classification of component faults in the MPS of different

Friction, adhesion and corrosion performance of metallurgical coatings in HTGR-helium

International Nuclear Information System (INIS)

Engel, R.; Kleemann, W.

1981-01-01

The friction-, adhesion-, thermal cycling- and corrosion performance of several metallurgical coating systems have been tested in a simulated HTGR-test atmosphere at elevated temperatures. The coatings were applied to a solid solution strengthened Ni-based superalloy. Component design requires coatings for the protection of mating surfaces, since under reactor operating conditions, contacting surfaces of metallic components under high pressures are prone to friction and wear damage. The coatings will have to protect the metal surface for 30 years up to 950 0 C in HTGR-helium. The materials tested were various refractory carbides with or without metallic binders and intermetallic compounds. The coatings evaluated were applied by plasma spraying-, detonation gun- and chemical vapor deposition techniques. These yielded two types of coatings which employ different mechanisms to improve the tribiological properties and maintain coating integrity. (Auth.)

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

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

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

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

International Nuclear Information System (INIS)

Xing Huiping; Xue Song

2014-01-01

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

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

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

Air ingress behavior during a primary-pipe rupture accident of HTGR

International Nuclear Information System (INIS)

Takeda, Tetsuaki

1997-11-01

The inherent properties of a HTGR facilitates the design with high degree of passive safe performances, compared to other type. However, it is still not clear if the present HTGR can maintain a passive safe function during a primary-pipe rupture accident, or what would be design criteria to guarantee the HTGR with the high degree of passive safe performances during the accident. To investigate safe characteristics, the study has been performed experimentally and analytically on the air ingress behavior during the accident. It was indicated that there are two stages in the accident of the HTGR having a reverse U-shaped channel. In the first stage, an air ingress process limits molecular diffusion and natural circulation of the gas mixture having a very slow velocity. In the second stage, the air ingress process limits the ordinary natural circulation of air throughout the reactor. A numerical calculation code has been developed to analyze thermal-hydraulic behavior during the first stage. This code provides a numerical method for analyzing a transport phenomena in a multi-component gas system by solving one-dimensional basic equations and using a flow network model. It was possible to predict or analyze the air ingress process regarding the density of the gas mixture, concentration of each gas species and duration of the first stage of the accident. It was indicated that the safe characteristics of the HTGR from the present experiment as follows. The safety cooling rate that the air ingress process terminates during the first stage exists in the HTGR having the reverse U-shaped channel. Moreover, the ordinary natural circulation of air can not produce in the second stage by injecting helium from the bottom of the pressure vessel corresponding the low-temperature side channel. Therefore, it was found that the idea of helium injection is one of useful methods for the prevention of air ingress and of graphite corrosion in the future HTGRs. (J.P.N.). 74 refs

2000 MW(t) HTGR-DC-GT Modesto Site dry cooled model 346 concice

International Nuclear Information System (INIS)

1979-07-01

Construction information is presented for a 800 MW(e) HTGR power reactor. The information is itemized for each reactor component or system and incudes quantity, labor hours, labor cost, material cost, and total costs

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

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

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

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.

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)

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

Creep-Rupture Properties and Corrosion Behaviour of 21/4 Cr-1 Mo Steel and Hastelloy X-Alloys in Simulated HTGR Environment

DEFF Research Database (Denmark)

Lystrup, Aage; Rittenhouse, P. L.; DiStefano, J. R.

Hastelloy X and 2/sup 1///sub 4/ Cr-1 Mo steel are being considered as structural alloys for components of a High-Temperature Gas-Cooled Reactor (HTGR) system. Among other mechanical properties, the creep behavior of these materials in HTGR primary coolant helium must be established to form part...

FY 1981 HTGR program summary-level program outline (revision 1/30/81)

International Nuclear Information System (INIS)

1981-01-01

The objective of the DOE HTGR Program is the development of technology for the most important HTGR applications. Through this support, DOE seeks to encourage private sector initiatives which will lead to the development of commercially attractive HTGR applications that concurrently support national energy goals. Currently perceived as important to national energy goals are applications that primarily address the process heat market with a view toward reduction of national requirements for oil, natural gas and coal. A high priority during FY 1981, therefore, will be to further identify and define the details of the Technology Program so as to assure that it is both necessary and sufficient to provide the required support. In the establishment of a supportive Technology Program, key elements which will be addressed are as follows: studies will be conducted to further identify and characterize important unique HTGR applications and to evaluate their potential in the context of market opportunities, utility/user interest, and national objectives to develop new energy supply options; based upon the configurations and operating characteristics projected for selected applications, Technology Program requirements must be identified to support development, verification, and ultimately licensing of components and systems comprising the facilities of interest; and in the context of limited resources, sufficient analysis and evaluation must be accomplished so as to prioritize technology elements in accordance with appropriately developed criteria

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

Feasibility of monitoring the strength of HTGR core support graphite: Part III

International Nuclear Information System (INIS)

Morgan, W.C.; Davis, T.J.; Thomas, M.T.

1983-02-01

Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR). The results reported herein pertain to the development of techniques for monitoring the core-support blocks; the PGX graphite used in these studies is the grade used for the core-support blocks of the Fort St. Vrain HTGR, and is coarser-grained than the grades used in our previous investigations. The through-transmission ultrasonic velocity technique, developed for monitoring strength of the core-support posts, is not suitable for use on the core-support blocks. Eddy-current and ultrasonic backscattering techniques have been shown to be capable of measuring the density-depth profile in oxidized PGX and, combined with a correlation of strength versus density, could yield an estimate of the strength-depth profile of in-service HTGR core support blocks. Correlations of strength versus density and other properties, and progress on the development of the eddy-current and ultrasonic backscattering techniques are reported

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

The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

2016-09-15

A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

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.

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

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

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)

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

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

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

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.)

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

HTGR-GT primary coolant transient resulting from postulated turbine deblading

International Nuclear Information System (INIS)

Cadwallader, G.J.; Deremer, R.K.

1980-11-01

The turbomachine is located within the primary coolant system of a nuclear closed cycle gas turbine plant (HTGR-GT). The deblading of the turbine can cause a rapid pressure equilibration transient that generates significant loads on other components in the system. Prediction of and design for this transient are important aspects of assuring the safety of the HTGR-GT. This paper describes the adaptation and use of the RATSAM program to analyze the rapid fluid transient throughout the primary coolant system during a spectrum of turbine deblading events. Included are discussions of (1) specific modifications and improvements to the basic RATSAM program, which is also briefly described; (2) typical results showing the expansion wave moving upstream from the debladed turbine through the primary coolant system; and (3) the effect on the transient results of different plenum volumes, flow resistances, times to deblade, and geometries that can choke the flow

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

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

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

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.

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)

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

International Nuclear Information System (INIS)

Dee, J.B.; Macken, T.

1977-01-01

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

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

HTGR plant availability and reliability evaluations. Volume I. Summary of evaluations

International Nuclear Information System (INIS)

Cadwallader, G.J.; Hannaman, G.W.; Jacobsen, F.K.; Stokely, R.J.

1976-12-01

The report (1) describes a reliability assessment methodology for systematically locating and correcting areas which may contribute to unavailability of new and uniquely designed components and systems, (2) illustrates the methodology by applying it to such components in a high-temperature gas-cooled reactor [Public Service Company of Colorado's Fort St. Vrain 330-MW(e) HTGR], and (3) compares the results of the assessment with actual experience. The methodology can be applied to any component or system; however, it is particularly valuable for assessments of components or systems which provide essential functions, or the failure or mishandling of which could result in relatively large economic losses

HTGR plant availability and reliability evaluations. Volume I. Summary of evaluations

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, G.J.; Hannaman, G.W.; Jacobsen, F.K.; Stokely, R.J.

1976-12-01

The report (1) describes a reliability assessment methodology for systematically locating and correcting areas which may contribute to unavailability of new and uniquely designed components and systems, (2) illustrates the methodology by applying it to such components in a high-temperature gas-cooled reactor (Public Service Company of Colorado's Fort St. Vrain 330-MW(e) HTGR), and (3) compares the results of the assessment with actual experience. The methodology can be applied to any component or system; however, it is particularly valuable for assessments of components or systems which provide essential functions, or the failure or mishandling of which could result in relatively large economic losses.

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

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)

Utilization of HTGR on active carbon recycling energy system

Energy Technology Data Exchange (ETDEWEB)

Kato, Yukitaka, E-mail: yukitaka@nr.titech.ac.jp

2014-05-01

A new energy transformation concept based on carbon recycling, called as active carbon recycling energy system, ACRES, was proposed for a zero carbon dioxide emission process. The ACRES is driven availably by carbon dioxide free primary energy. High temperature gas cooled reactor (HTGR) is a candidate of the energy sources for ACRES. A smart ironmaking system with ACRES (iACRES) is one of application examples. The contribution of HTGR on iACRES was discussed thermodynamically in this study. A carbon material is re-used cyclically as energy carrier media in ACRES. Carbon monoxide (CO) had higher energy densities than hydrogen and was compatible with conventional process. Thus, CO was suitable recycling media for ACRES. Efficient regeneration of CO was a key technology for ACRES. A combined system of hydrogen production by water electrolysis and CO{sub 2} hydrogen reduction was candidate. CO{sub 2} direct electrolysis was also one of the candidates. HTGR was appropriate heat source for both water and CO{sub 2} electrolysises, and CO{sub 2} hydrogen reduction. Thermodynamic energy balances were calculated for both systems with HTGR for an ironmaking system. The direct system showed relatively advantage to the combined system in the stand point of enthalpy efficiency and simplicity of the process. One or two plants of HTGR are corresponding with ACRES system for one unit of conventional blast furnace. The proposed ACRES system with HTGR was expected to form the basis of a new energy industrial process that had low CO{sub 2} emission.

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

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)

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

Design of the HTGR for process heat applications

International Nuclear Information System (INIS)

Vrable, D.L.; Quade, R.N.

1980-05-01

This paper discusses a design study of an advanced 842-MW(t) HTGR with a reactor outlet temperature of 850 0 C (1562 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

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

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

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

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

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)

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

Public acceptance of HTGR technology - HTR2008-58218

International Nuclear Information System (INIS)

Hannink, R.; Kuhr, R.; Morris, T.

2008-01-01

Nuclear energy projects continue to evoke strong emotional responses from the general public throughout the world. High Temperature Gas-Cooled Reactor (HTGR) technology offers improved safety and performance characteristics that should enhance public acceptance but is burdened with demonstrating a different set of safety principles. This paper summarizes key issues impacting public acceptance and discusses the importance of openly engaging the public in the early stages of new HTGR projects. The public gets information about new technologies through schools and universities, news and entertainment media, the internet, and other forms of information exchange. Development of open public forums, access to information in understandable formats, participation of universities in preparing and distributing educational materials, and other measures will be needed to support widespread public confidence in the improved safety and performance characteristics of HTGR technology. This confidence will become more important as real projects evolve and participants from outside the nuclear industry begin to evaluate the real and perceived risks, including potential impacts on public relations, branding, and shareholder value when projects are announced. Public acceptance and support will rely on an informed understanding of the issues and benefits associated with HTGR technology. Major issues of public concern include nuclear safety, avoidance of greenhouse gas emissions, depletion of natural gas resources, energy security, nuclear waste management, local employment and economic development, energy prices, and nuclear proliferation. Universities, the media, private industry, government entities, and other organizations will all have roles that impact public acceptance, which will likely play a critical role in the future markets, siting, and permitting of HTGR projects. (authors)

HTGR structural-materials efforts in the US

International Nuclear Information System (INIS)

Rittenhouse, P.L.; Roberts, D.I.

1982-07-01

The status of ongoing structural materials programs being conducted in the US to support development and deployment of the high-temperature gas-cooled reactor (HTGR) is described. While the total US program includes work in support of all variants of this reactor system, the emphasis of this paper is on the work aimed at support of the steam cycle/cogeneration (SC/C) version of the HTGR. Work described includes activities to develop design and performance prediction data on metals, ceramics, and graphite

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

Analysis of some accident conditions in confirmation of the HTGR safety

Energy Technology Data Exchange (ETDEWEB)

Grebennik, V. N.; Grishanin, E. I.; Kukharkin, N. E.; Mikhailov, P. V.; Pinchuk, V. V.; Ponomarev-Stepnoy, N. N.; Fedin, G. I.; Shilov, V. N.; Yanushevich, I. V. [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

1981-01-15

This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved.

Analysis of some accident conditions in confirmation of the HTGR safety

International Nuclear Information System (INIS)

Grebennik, V.N.; Grishanin, E.I.; Kukharkin, N.E.; Mikhailov, P.V.; Pinchuk, V.V.; Ponomarev-Stepnoy, N.N.; Fedin, G.I.; Shilov, V.N.; Yanushevich, I.V.

1981-01-01

This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved

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

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

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

Survey on the activities in Switzerland in the field of HTGR-development

International Nuclear Information System (INIS)

Sarlos, G.; Brogli, R.; Mathews, D.; Bucher, K.H.; Helbling, W.

1991-01-01

The activities of the Swiss industry and of the ''Paul Scherrer Institute'' in the development and production of components and systems for the nuclear industry are reviewed. For the HTGR, major programs include the German HTR-500 project, the gas-cooled district heating reactor (GHR), and the PROTEUS critical experiments. The experiments are being performed in the framework of an IAEA coordinated research program. (author)

HTGR-GT and electrical load integrated control

International Nuclear Information System (INIS)

Chan, T.; Openshaw, F.; Pfremmer, D.

1980-05-01

A discussion of the control and operation of the HTGR-GT power plant is presented in terms of its closely coupled electrical load and core cooling functions. The system and its controls are briefly described and comparisons are made with more conventional plants. The results of analyses of selected transients are presented to illustrate the operation and control of the HTGR-GT. The events presented were specifically chosen to show the controllability of the plant and to highlight some of the unique characteristics inherent in this multiloop closed-cycle plant

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)

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

CONTEMPT-G computer program and its application to HTGR containments

International Nuclear Information System (INIS)

Macnab, D.I.

1976-03-01

The CONTEMPT-G computer program has been developed by General Atomic Company to simulate the temperature-pressure response of a containment atmosphere to postulated depressurization of High-Temperature Gas-Cooled Reactor (HTGR) primary or secondary coolant circuits. The mathematical models currently used in the code are described, and applications of the code in examples of the atmospheric response of a representative containment to a variety of postulated HTGR accident conditions are presented. In particular, maximum containment temperature and pressure, equilibrated long-term prestressed concrete reactor vessel and containment pressures, and peak containment conditions following steam pipe ruptures are examined for a representative 770-MW(e) HTGR

HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

International Nuclear Information System (INIS)

Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin

2011-01-01

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

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

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)

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)

Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

International Nuclear Information System (INIS)

Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

2005-01-01

Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

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.

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)

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

SC-HTGR Performance Impact for Arid Sites

International Nuclear Information System (INIS)

Lommers, L.; Geschwindt, J.; Southworth, F.; Shahrokhi, F.

2014-01-01

The SC-HTGR provides high temperature steam which can support industrial process heat applications as well as high efficiency electricity generation. The increased generating efficiency resulting from using high steam temperature provides greater plant output than lower temperature concepts, and it also reduces the fraction of waste heat which must be rejected. This capability is particularly attractive for sites with little or no water for heat rejection. This high temperature capability provides greater flexibility for these sites, and it results in a smaller performance penalty than for lower temperature systems when dry cooling must be used. The performance of the SC-HTGR for a conventional site with wet cooling is discussed first. Then the performance for arid sites is evaluated. Dry cooling performance is evaluated for both moderately arid sites and very hot sites. Offdesign performance of the dry cooling system under extreme conditions is also considered. Finally, operating strategies are explored for sites where some cooling water may be available but only in very limited quantities. Results of these assessments confirm that the higher operating temperatures of the SC-HTGR are very beneficial for arid sites, providing significant advantages for both gross and net power generation. (author)

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

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)

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)

Application of the lines-of-protection concept to the HTGR-SC/C

International Nuclear Information System (INIS)

1981-09-01

The purpose of this document is to present a method for structuring the safety related design and development plans for the HTGR. This method centers on and develops the concept that the HTGR inherently (and by design) provides independent and successive LOPs against potential core related accidents and any resulting public harm. To exemplify the LOP concept and its application to the HTGR, this document identifies some key bases and assumptions, describes the four LOPs selected for the HTGR, identifies the associated safety goals and plant success criteria, and establishes methods for safety research and development prioritization. A task breakdown structure is then described, which in a complete hierarchial fashion can be used to catalog all safety related tasks necessary to demonstrate LOP success as well as catalog safety research areas which cannot be conveniently grouped under the LOPs

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

Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys

International Nuclear Information System (INIS)

Aronson, S.; Chow, J.G.Y.; Soo, P.; Friedlander, M.

1978-01-01

Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650 0 C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTGR materials to more prototypic, low concentrations of I 2 , Te 2 and CsI in the presence of low partial pressures of O 2 . These samples are scheduled for mechanical bend tests after exposure to determine the degree of embrittlement

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)

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

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

International Nuclear Information System (INIS)

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

1989-04-01

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

Basic principles on the safety evaluation of the HTGR hydrogen production system

International Nuclear Information System (INIS)

Ohashi, Kazutaka; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko

2009-03-01

As HTGR hydrogen production systems, such as HTTR-IS system or GTHTR300C currently being developed by Japan Atomic Energy Agency, consists of nuclear reactor and chemical plant, which are without a precedent in the world, safety design philosophy and regulatory framework should be newly developed. In this report, phenomena to be considered and events to be postulated in the safety evaluation of the HTGR hydrogen production systems were investigated and basic principles to establish acceptance criteria for the explosion and toxic gas release accidents were provided. Especially for the explosion accident, quantitative criteria to the reactor building are proposed with relating sample calculation results. It is necessary to treat abnormal events occurred in the hydrogen production system as an 'external events to the nuclear plant' in order to classify the hydrogen production system as no-nuclear facility' and basic policy to meet such requirement was also provided. (author)

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)

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

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

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

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

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)

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

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.

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.

Generation of a Broad-Group HTGR Library for Use with SCALE

International Nuclear Information System (INIS)

Ellis, Ronald James; Lee, Deokjung; Wiarda, Dorothea; Williams, Mark L.; Mertyurek, Ugur

2012-01-01

With current and ongoing interest in high temperature gas reactors (HTGRs), the U.S. Nuclear Regulatory Commission (NRC) anticipates the need for nuclear data libraries appropriate for use in applications for modeling, assessing, and analyzing HTGR reactor physics and operating behavior. The objective of this work was to develop a broad-group library suitable for production analyses with SCALE for HTGR applications. Several interim libraries were generated from SCALE fine-group 238- and 999-group libraries, and the final broad-group library was created from Evaluated Nuclear Data File/B Version ENDF/B-VII Release 0 cross-section evaluations using new ORNL methodologies with AMPX, SCALE, and other codes. Furthermore, intermediate resonance (IR) methods were applied to the HTGR broadgroup library, and lambda factors and f-factors were incorporated into the library s nuclear data files. A new version of the SCALE BONAMI module named BONAMI-IR was developed to process the IR data in the new library and, thus, eliminate the need for the CENTRM/PMC modules for resonance selfshielding. This report documents the development of the HTGR broad-group nuclear data library and the results of test and benchmark calculations using the new library with SCALE. The 81-group library is shown to model HTGR cases with similar accuracy to the SCALE 238-group library but with significantly faster computational times due to the reduced number of energy groups and the use of BONAMI-IR instead of BONAMI/CENTRM/PMC for resonance self-shielding calculations.

Present Status of HTGR Utilization System Development in Japan

International Nuclear Information System (INIS)

Miyamoto, Yoshiaki

2000-01-01

Efforts are to be continuously devoted to establish and upgrade HTGR technology in the world. Japan Atomic Energy Research Institute (JAERI) has conducted the R and D of HTGRs since the 1960's in Japan, focusing on mainly the construction of High Temperature engineering Test Reactor (HTTR) which is an HTGR with a maximum helium gas temperature of 950 o C at the reactor outlet and HTGR utilization systems. The HTTR achieved first criticality on November 10, 1998 and will restart from January in 2001. In the R and D program of HTGR utilization systems, JAERI has conducted hydrogen production systems with HTGR to demonstrate the applicability of nuclear heat for extensive energy demands besides the electric power generation. JAERI has developed a hydrogen production system by steam reforming process of natural gas using nuclear heat supplied from the HTTR. Prior to the demonstration test of HTTR hydrogen production system, a 1/30-scale out-of-pile test facility is under construction for safety review and detailed design of the system. The out-of-pile test facility will be started in 2001 and will be continued about 4 years. The hydrogen permeation and corrosion tests have been carried out since 1997. Check and review for the demonstration program in the HTTR hydrogen production system will be made in 2001. Then the HTTR hydrogen production system is scheduled to be constructed from 2003 and demonstratively operated from around 2006. In parallel with the R and D of the HTTR hydrogen production system, hydrogen production method by thermochemical water splitting, so-called IS process, has been studied in JAERI. The IS process is placed as one of future candidates of the heat utilization systems of the HTTR following the steam reforming system. Continuous and stoichiometric production of hydrogen and oxygen for 48 hours was successfully achieved with a laboratory-scale apparatus mainly made of glass. Following this achievement, the study has been continued with a larger

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

The effect of creep-fatigue damage relationships upon HTGR heat exchanger design

International Nuclear Information System (INIS)

Kozina, M.M.; King, J.H.; Basol, M.

1984-01-01

Mo tubes followed by a superheater of straight tubes of Alloy 800H in the central zone of the steam generator. The high-temperature components affected by creep-fatigue interaction are the tubing and the superheated steam tubesheet of Alloy 800H. The effects of the revised creep-fatigue damage relationships were evaluated by: (1) calculating the temperature-dependent allowable strain range with the assumed bilinear damage relationship for Alloy 800H; (2) calculating the temperature-dependent allowable strain range with reduced fatigue allowables for 2-1/4 Cr-1 Mo; and (3) predicting the strain range in the critical parts by extrapolation of finite element calculations for the second or last cycle analyzed for each transient to the expected number of cycles and hold times. The preliminary analyses indicate that the Alloy 800H tubing and tubesheets have predicted strains substantially under the allowables based upon the bilinear damage relationship but that the 2-1/4 Cr-1 Mo tubing at the hot end inner radius portion of the EES tube bundle presents a slightly overstressed situation. It is believed that there is sufficient design latitude to resolve this problem in the continuing preliminary design. It is concluded that the HTGR-SC/C steam generator design has sufficient margin to accommodate the more conservative creep-fatigue damage relationships. (author)

HTGR core seismic analysis using an array processor

International Nuclear Information System (INIS)

Shatoff, H.; Charman, C.M.

1983-01-01

A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

HTGR technology development in Japan advances so much. Leading world technology to global standards

International Nuclear Information System (INIS)

Ogawa, Masuro; Hino, Ryutaro; Kunitomi, Kazuhiko; Onuki, Kaoru; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Sawa, Kazuhiro

2007-01-01

The JAEA has conducted research and development of HTGR for hydrogen production since 1969 and attained the operation of 950degC at reactor coolant outlet of the HTTR in 2004. This article describes present status and future plan of R and D in the area of HTGR technology and high temperature heat utilization and also introduces the design of the commercial HTGR cogeneration system based on R and D results leading to world standards. (T. Tanaka)

Status of the HTGR development program in Japan

International Nuclear Information System (INIS)

Saito, S.

1991-01-01

According to the revision of the Long-Term Program for Development and Utilization of Nuclear Energy issued by the Japanese Atomic Energy Commission, High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, will be constructed by the Japan Atomic Energy Research Institute (JAERI) in order to establish and upgrade the technology basis for an HTGR, serving at the same time as a potential tool for new and innovative basic research. The budget for the construction of the HTTR was approved by the Government and JAERI is now proceeding with the construction design of the HTTR, focussing the first criticality in the end of FY 1995. In order to establish and upgrade HTGR technology basis systematically and efficiently, and also to carry out innovative basic research on high temperature technologies, Japan will perform necessary R and D mainly at JAERI, which is a leading organization of the R and D. In addition, in order to promote the R and D on HTGRs more efficiently, Japan will promote the existing international cooperation with the research organizations in foreign countries. (author). 5 figs, 3 tabs

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)

An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

2008-03-01

To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

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

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

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)

Universally applicable design concept of stably controlling an HTGR-hydrogen production system

International Nuclear Information System (INIS)

Hada, Kazuhiko; Shibata, Taiju; Nishihara, Tetsuo; Shiozawa, Shusaku

1996-01-01

An HTGR-hydrogen production system should be designed to have stable controllability because of a large difference in thermal dynamics between reactor and hydrogen production system and such a control design concept should be universally applicable to a variety of hydrogen production processes by the use of nuclear heat from HTGR. A transient response analysis of an HTGR-steam reforming hydrogen production system showed that a steam generator installed in a helium circuit for cooling the nuclear reactor provides stable controllability of the total system, resulting in avoiding a reactor scram. A survey of control design-related characteristics among several hydrogen production processes revealed the similarity of endothermic chemical reactions by the use of high temperature heat and that steam is required as a reactant of the endothermic reaction or for preheating a reactant. Based on these findings, a system design concept with stable controllability and universal applicability was proposed to install a steam generator as a downstream cooler of an endothermic reactor in the helium circuit of an HTGR-hydrogen production system. (author)

Status, results and usefulness of risk analyses for HTGR type reactors of different capacity accessory to planning

International Nuclear Information System (INIS)

Kroeger, W.; Mertens, J.

1985-01-01

As regards system-inherent risks, HTGR type reactors are evaluated with reference to the established light-water-moderated reactor types. Probabilistic HTGR risk analyses have shown modern HTGR systems to possess a balanced safety concept with a risk remaining distinctly below legally accepted values. Inversely, the development and optimization of the safety concepts have been (and are being) essentially co-determined by the probabilistic analyses, as it is technically sensible and economically necessary to render the specific safety-related HTGR properties eligible for licensing. (orig./HP) [de

HTGR gas turbine program. Semiannual progress report, April 1-September 30, 1978

International Nuclear Information System (INIS)

1979-12-01

This report describes work performed under the gas turbine HTGR (HTGR-GT) program, Department of Energy Contract DE-AT03-76-SF70046, during the period April 1, 1978 through September 30, 1978. The work reported covers the demonstration and commercial plant concept studies including plant layout, heat exchanger studies, turbomachine studies, systems analysis, and reactor core engineering

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.)

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.)

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)

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)

Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

Energy Technology Data Exchange (ETDEWEB)

Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2017-05-15

Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

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

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

HTGR containment design options: an application of probabilistic risk assessment

International Nuclear Information System (INIS)

1977-08-01

Through the use of probabilistic risk assessment (PRA), it is possible to quantitatively evaluate the radiological risk associated with a given reactor design and to place such risk into perspective with alternative designs. The merits are discussed for several containment alternatives for the HTGR from the viewpoints of economics and licensability, as well as public risk. The quantification of cost savings and public risk indicates that presently acceptable public risk can be maintained and cost savings of $40 million can result from use of a vented confinement for the HTGR

GTOROTO: a simulation system for HTGR core seismic behavior

International Nuclear Information System (INIS)

Ikushima, Takeshi; Nakamura, Yasuhiro; Onuma, Yoshio

1980-07-01

One of the most important design of HTGR core is its aseismic structure. Therefore, it is necessary to predict the forces and motion of the core blocks. To meet the requirement, many efforts to develop analytical methods and computer programs are made. A graphic simulation system GTOROTO with a CRT graphic display and lightpen was developed to analyze the HTGR core behavior in seismic excitation. Feature of the GTOROTO are as follows: (1) Behavior of the block-type HTGR core during earthquake can be shown on the CRT-display. (2) Parameters of the computing scheme can be changed with the lightpen. (3) Routines of the computing scheme can be changed with the lightpen and an alteration switch. (4) Simulation pictures are shown automatically. Hardcopies are available by plotter in stopping the progress of simulation pictures. Graphic representation can be re-start with the predetermined program. (5) Graphic representation informations can be stored in assembly language on a disk for rapid representation. (6) A computer-generated cinema can be made by COM (Computer Output Microfilming) or filming directly the CRT pictures. These features in the GTOROTO are provided in on-line conversational mode. (author)

Analysis of fission product release from HTGR core during transient temperature excursion

International Nuclear Information System (INIS)

Saito, Takao; Yamatoya, Naotoshi; Onuma, Mamoru

1978-01-01

The computer program ''FRANC'' was developed to calculate the release activity of fission products from a high-temperature gas cooled reactor (HTGR) core during transient temperature excursions such as a hypothetical loss of forced circulation combined with design basis depressurization. The program utilizes a segmented cylindrical core spatial model with the associated values of the prior fuel irradiation history and temperature conditions. The fission product transport and decay chain behavior is expressed by a set of differential equations. This set of equations describes the entire core inventory of fission products by means of calculated parameters based on the detailed spatial core conditions. The program determines the time-dependent amounts of fission product nuclides escaping from the core into the coolant. Coded in Continuous System Simulation Language (CSSL) with double precision, FRANC showed appropriate results for both short- and long-lived fission product nuclides. The sample calculation conducted by applying the program to a large HTGR indicated that it would take about one hour for noble gases and volatile nuclides to be released to the coolant, and several hours for metalic nuclides. (auth.)

Reduced risk HTGR concept for industrial heat application

International Nuclear Information System (INIS)

Boardman, C.E.; Lipps, A.J.

1982-01-01

The industrial process heat market has been identified as major market for the High Temperature Gas-Cooled Reactor (HTGR), however, this market introduces stringent availability requirements on the reactor system relative to electric plants which feed a large existing grid. The characteristics and requirements of the industrial heat markets are summarized; the risks associated with serving this market with a single large HTGR will be discussed; and the modular concept, which has the potential to reduce both safety and investment risks, will be described. The reference modular concept described consists of several small, relatively benign nuclear heat sources linked together to supply heat energy to a balance-of-plant incorporating a process gas train/thermochemical pipe line system and a normal steam-electric plant

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

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

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)

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

1170-MW(t) HTGR-PS/C plant application-study report: alumina-plant application

International Nuclear Information System (INIS)

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

1981-05-01

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

ORNL's NRC-sponsored HTGR safety and licensing analysis activities for Fort St. Vrain and advanced reactors

International Nuclear Information System (INIS)

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

1985-01-01

The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. The program has also emphasized code verification, using Fort St. Vrain data where applicable, and comparing results with industry-generated codes. By the use of model and parameter adjustment routines, safety-significant uncertainties have been identified. A major part of the analysis work has been done for the Fort St. Vrain HTGR, and has included analyses of FSAR accident scenario re-evaluations, the core block oscillation problem, core support thermal stress questions, technical specification upgrade review, and TMI action plan applicability studies. The large, 2240-MW(t) cogeneration lead plant design was analyzed in a multi-laboratory cooperative effort to estimate fission product source terms from postulated severe accidents

Overview of HTGR utilization system developments at JAERI

International Nuclear Information System (INIS)

Miyamoto, Y.; Shiozawa, S.; Inagaki, Y.

1997-01-01

JAERI has been constructing a 30-MWt HTGR, named HTTR, to develop technology and to demonstrate effectiveness of high-temperature nuclear heat utilization. A hydrogen production system by natural gas steam reforming is to be the first heat utilization system of the HTTR since its technology matured in fossil-fired plant enables to couple with HTTR in the early 2000's and technical solutions demonstrated by the coupling will contribute to all other hydrogen production systems. The HTTR steam reforming system is designed to utilize the nuclear heat effectively and to achieve hydrogen productivity competitive to that of a fossil-fired plant with operability, controllability and safety acceptable enough to commercialization, and an arrangement of key components was already decided. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test is planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions. The out-of-pile system is an approximately 1/20-1/30 scale system of the HTTR steam reforming system and simulates key components downstream from an IHX

Feasibility of monitoring the strength of HTGR core support graphite. Part II

International Nuclear Information System (INIS)

Morgan, W.C.; Becker, F.L.

1979-08-01

The results reported establish the technical feasibility of a method for monitoring the strength of HTGR core support structures in situ. Correlations have been established between the velocity of an ultrasonic pulse and the compressive strength of four different grades of graphite. For some grades of graphite, one or more of the correlations are practically independent of oxidation profile in samples having cylindrical geometry (as in the core support posts). For other grades of graphite, and for other sample geometries, the oxidation-depth profile must be known in order to reliably predict the effect of oxidation on compressive strength

The desorption of caesium from Peach Bottom HTGR steam generator materials

International Nuclear Information System (INIS)

Clark, M.J.

1979-03-01

The work at Harwell on the Peach Bottom End-of-Life Program in co-operation with the General Atomic Company (U.S.A.) is described. Materials taken from the Economiser, Evaporator and Superheater Sections of the Peach Bottom Unit No. 1. High Temperature Gas Cooled Reactor (HTGR) Heat Exchanger were placed in a reducing atmosphere comparable to the composition of an HTGR helium coolant gas, and the desorption of caesium isotopes measured under known conditions of flow, temperature and oxygen pressure. (author)

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

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)

Dynamics and control modeling of the closed-cycle gas turbine (GT-HTGR) power plant

International Nuclear Information System (INIS)

Bardia, A.

1980-02-01

The simulation if presented for the 800-MW(e) two-loop GT-HTGR plant design with the REALY2 transient analysis computer code, and the modeling of control strategies called for by the inherently unique operational requirements of a multiple loop GT-HTGR is described. Plant control of the GT-HTGR is constrained by the nature of its power conversion loops (PCLs) in which the core cooling flow and the turbine flow are directly related and thus changes in flow affect core cooling as well as turbine power. Additionally, the high thermal inertia of the reactor core precludes rapid changes in the temperature of the turbine inlet flow

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.)

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

Examination on small-sized cogeneration HTGR for developing countries

International Nuclear Information System (INIS)

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; Mozumi, Yasuhiro; Imai, Yoshiyuki; Tanaka, Nobuyuki; Okuda, Hiroyuki; Iwatsuki, Jin; Kubo, Shinji; Takada, Shoji; Nishihara, Tetsuo; Kunitomi, Kazuhiko

2008-03-01

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors. (author)

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)

Draft of standard for graphite core components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

2010-01-01

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

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

Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

International Nuclear Information System (INIS)

Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

2014-01-01

Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

1170-MW(t) HTGR-PS/C plant application study report: heavy oil recovery application

International Nuclear Information System (INIS)

Rao, R.; McMain, A.T. Jr.

1981-05-01

This report describes the application of a high-temperature gas-cooled reactor (HTGR) which operates in a process steam/cogeneration (PS/C) mode in supplying steam for enhanced recovery of heavy oil and in exporting electricity. The technical and economic merits of an 1170-MW(t) HTGR-PS/C are compared with those of coal-fired plants and (product) oil-fired boilers for this application. The utility requirements for enhanced oil recovery were calculated by establishing a typical pattern of injection wells and production wells for an oil field similar to that of Kern County, California. The safety and licensing issues of the nuclear plant were reviewed, and a comparative assessment of the alternative energy sources was performed. Technically and economically, the HTGR-PS/C plant has attractive merits. The major offsetting factors would be a large-scale development of a heavy oil field by a potential user for the deployment of a 1170-MW(t) HTGR-PS/C; plant and the likelihood of available prime heavy oil fields for the mid-1990 operation

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.)

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)

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

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

Development of structural design procedure of plate-fin heat exchanger for HTGR

Energy Technology Data Exchange (ETDEWEB)

Mizokami, Yorikata, E-mail: yorikata_mizokami@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-Chome, Hyogo-ku, Kobe 652-8585 (Japan); Igari, Toshihide [Mitsubishi Heavy Industries, Ltd., 5-717-1, Fukahori-machi, Nagasaki 851-0392 (Japan); Kawashima, Fumiko [Kumamoto University, 39-1 Kurokami 2-Chome, Kumamoto 860-8555 (Japan); Sakakibara, Noriyuki [Mitsubishi Heavy Industries, Ltd., 5-717-1, Fukahori-machi, Nagasaki 851-0392 (Japan); Tanihira, Masanori [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo 108-8215 (Japan); Yuhara, Tetsuo [The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Hiroe, Tetsuyuki [Kumamoto University, 39-1 Kurokami 2-Chome, Kumamoto 860-8555 (Japan)

2013-02-15

Highlights: ► We propose high temperature structural design procedure for plate-fin heat exchanger ► Allowable stresses for brazed structures will be newly discussed ► Validity of design procedure is confirmed by carrying out partial model tests ► Proposed design procedure is applied to heat exchangers for HTGR. -- Abstract: Highly efficient plate-fin heat exchanger for application to HTGR has been focused on recently. Since this heat exchanger is fabricated by brazing a lot of plates and fins, a new procedure for structural design of brazed structures in the HTGR temperature region up to 950 °C is required. Firstly in this paper influences on material strength due to both thermal aging during brazing process and helium gas environment were experimentally examined, and failure mode and failure limit of brazed side-bar structures were experimentally clarified. Secondly allowable stresses for aging materials and brazed structures were newly determined on the basis of the experimental results. For the purpose of validating the structural design procedure including homogenization FEM modeling, a pressure burst test and a thermal fatigue test of partial model for plate-fin heat exchanger were carried out. Finally, results of reference design of plate-fin heat exchangers of recuperator and intermediate heat exchanger for HTGR plant were evaluated by the proposed design criteria.

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)

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

Development of seismic analysis model for HTGR core on commercial FEM code

International Nuclear Information System (INIS)

Tsuji, Nobumasa; Ohashi, Kazutaka

2015-01-01

The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

New HTGR plant concept with inherently safe features aimed at small energy users needs

International Nuclear Information System (INIS)

McDonald, C.F.; Silady, F.S.; Shenoy, A.S.

1982-01-01

A small high-temperature gas-cooled reactor (HTGR) concept is proposed which could provide the energy needs for certain sectors of industrialized nations and the developing countries. The key to the economic success for small reactors, which have potential benefits for special markets, lies in altering the traditional scaling laws. Toward this goal, a small HTGR concept embodying passive decay heat removal features is currently being evaluated. This paper emphasizes the safety-related aspects of a small HTGR. The proposed small reactor concept is new and still in the design development stage, and a significant effort must be expended to establish a design which is technically and economically feasible and will meet the increasingly demanding safety and licensing goals for reactors of the future

Further HTGR core support structure reliability studies. Interim report No. 1

International Nuclear Information System (INIS)

Platus, D.L.

1976-01-01

Results of a continuing effort to investigate high temperature gas cooled reactor (HTGR) core support structure reliability are described. Graphite material and core support structure component physical, mechanical and strength properties required for the reliability analysis are identified. Also described are experimental and associated analytical techniques for determining the required properties, a procedure for determining number of tests required, properties that might be monitored by special surveillance of the core support structure to improve reliability predictions, and recommendations for further studies. Emphasis in the study is directed towards developing a basic understanding of graphite failure and strength degradation mechanisms; and validating analytical methods for predicting strength and strength degradation from basic material properties

Recent activities on the HTGR for its commercialization in the 21st century

International Nuclear Information System (INIS)

Minatsuki, I.; Uchida, S.; Nomura, S.; Yamada, S.

1997-01-01

Currently, the greatest concern about energy is the need to rapidly increase the energy supply, while also conserving energy reserves and protecting the worldwide environment in the coming century. Furthermore, the direct use of thermal energy from nuclear reactors is an effective way to widen the application of nuclear energy. From this standpoint, Mitsubishi Heavy Industries (MHI) has been continuing the various activities related to the High Temperature Gas Cooled Reactor (HTGR). At present, MHI is participating in the High Temperature Engineering Test Reactor (HTTR) project, which is under construction at Oarai promoted by the Japan Atomic Energy Research Institute, as the primary fabricator. Moreover MHI has been conducting research and development to investigate the feasibility of HTGR commercialization in future. In this paper, the results of various studies are summarized to introduce our HTGR activities

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

The coupled code system TORT-TD/ATTICA3D for 3-D transient analysis of pebble-bed HTGR

International Nuclear Information System (INIS)

Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Laurien, E.; Bader, J.; EnBW Kernkraft GmbH, Philippsburg

2012-01-01

This paper describes the time-dependent 3-D discrete-ordinates based coupled code system TORT-TD/ATTICA3D and its application to HTGR of pebble bed type. TORT-TD/ATTICA3D is represented by a single executable and adapts the so-called internal coupling approach. Three-dimensional distributions of temperatures from ATTICA3D and power density from TORT-TD are efficiently exchanged by direct memory access of array elements via interface routines. Applications of TORT-TD/ATTICA3D to three transients based on the PBMR-400 benchmark (total and partial control rod withdrawal and cold helium ingress) and the full power steady state of the HTR-10 are presented. For the partial control rod withdrawal, 3-D effects of local neutron flux redistributions are clearly identified. The results are very promising and demonstrate that the coupled code system TORT-TD/ATTICA3D may represent a key component in a future comprehensive 3-D code system for HTGR of pebble bed type. (orig.)

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)

Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

International Nuclear Information System (INIS)

McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

1980-02-01

Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed

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

Safety analysis of coupling system of hybrid (MED-RO) nuclear desalination system utilising waste heat from HTGR

International Nuclear Information System (INIS)

Raha, Abhijit; Kishore, G.; Rao, I.S.; Adak, A.K.; Srivastava, V.K.; Prabhakar, S.; Tewari, P.K.

2010-01-01

To meet the generation IV goals, High Temperature Gas Cooled Reactors (HTGRs) are designed to have relatively higher thermal efficiency and enhanced safety and environmental characteristics. It can provide energy for combined production of hydrogen, electricity and other industrial applications. The waste heat available in the HTGR power cycle can also be utilized for the desalination of seawater for producing potable water. Desalination is an energy intensive process, so use of waste heat from HTGR certainly makes desalination process more affordable to create fresh water resources. So design of the coupling system, as per the safety design requirement of nuclear desalination plant, of desalination plant with HTGR is very crucial. In the first part of this paper, design of the coupling system between hybrid Multi Effect Desalination-Reverse Osmosis (MED-RO) nuclear desalination plant and HTGR to utilize the waste heat in HTGR are discussed. In the next part deterministic safety analysis of the designed coupling system of are presented in detail. It was found that all the coupling system meets the acceptance criteria for all the Postulated Initiating Events (PIE's) limited to DBA. (author)

Proceedings of the 1st JAEA/KAERI information exchange meeting on HTGR and nuclear hydrogen technology

International Nuclear Information System (INIS)

Sato, Hiroyuki; Sakaba, Nariaki; Nishihara, Tetsuo; Yan, Xing L.; Hino, Ryutaro

2007-03-01

Japan Atomic Energy Agency (JAEA) has completed an implementation with Korea Atomic Energy Research Institute (KAERI) on HTGR and nuclear hydrogen technology, 'The Implementation of Cooperative Program in the Field of Peaceful Uses of Nuclear Energy between KAERI and JAEA. 'To facilitate efficient technology development on HTGR and nuclear hydrogen by the IS process, an information exchange meeting was held at the Oarai Research and Development Center of JAEA on August 28-30, 2006 under Program 13th of the JAEA/KAERI Implementation, 'Development of HTGR and Nuclear Hydrogen Technology'. JAEA and KAERI mutually showed the status and future plan of the HTTR (High-Temperature Engineering Test Reactor) project in Japan and of the NHDD (Nuclear Hydrogen Development and Demonstration) project in Korea, respectively, and discussed collaboration items. This proceedings summarizes all materials of presented technical discussions on HTGR and hydrogen production technology as well as the meeting briefing including collaboration items. (author)

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

Study on Off-Design Steady State Performances of Helium Gas Turbo-compressor for HTGR-GT

International Nuclear Information System (INIS)

Qisen Ren; Xiaoyong Yang; Zhiyong Huang; Jie Wang

2006-01-01

The high temperature gas-cooled reactor (HTGR) coupled with direct gas turbine cycle is a promising concept in the future of nuclear power development. Both helium gas turbine and compressor are key components in the cycle. Under normal conditions, the mode of power adjustment is to control total helium mass in the primary loop using gas storage vessels. Meanwhile, thermal power of reactor core is regulated. This article analyzes off-design performances of helium gas turbine and compressors for high temperature gas-cooled reactor with gas turbine cycle (HTGR-GT) at steady state level of electric power adjustment. Moreover, performances of the cycle were simply discussed. Results show that the expansion ratio of turbine decreases as electric power reduces but the compression ratios of compressors increase, efficiencies of both turbine and compressors decrease to some extent. Thermal power does not vary consistently with electric power, the difference between these two powers increases as electric power reduces. As a result of much thermal energy dissipated in the temperature modulator set at core inlet, thermal efficiency of the cycle has a widely reduction under partial load conditions. (authors)

SONATINA-2V: a computer program for seismic analysis of the two-dimensional vertical slice HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi

1982-07-01

A computer program SONATINA-2V has been developed for predicting the behavior of a two-dimensional vertical slice HTGR core under seismic excitation. SONATINA-2V is a general two-dimensional computer program capable of analyzing the vertical slice HTGR core with the permanent side reflector blocks and its restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Coulomb friction is taken into account between blocks and between dowel pin and hole. A spring dashpot model is used for the collision process between adjacent blocks. The core support structure is represented by a single block. The computer program SONATINA-2V is capable of analyzing the core behavior for an excitation input applied simultaneously to both vertical and horizontal directions. Analytical results obtained from SONATINA-2V are compared with experimental results and are found to be in good agreement. The computer program can thus be used to predict with a good accuracy the behavior of the HTGR core under seismic excitation. In the present report are given, the theoretical formulation of the analytical model, a user's manual to describe the input and output format, and sample problems. (author)

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

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

Investigations of postulated accident sequences for the Fort St. Vrain HTGR

International Nuclear Information System (INIS)

Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M.; Sanders, J.P.

1978-01-01

The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models

The HTTR project as the world leader of HTGR research and development

International Nuclear Information System (INIS)

Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

2005-01-01

As a next generation type nuclear system which will expand nuclear energy use area with high temperature nuclear heat utilization and improve economic competitiveness greatly, High Temperature Gas-cooled Reactor (HTGR) has become the R and D item of prime importance at home as well as abroad to establish hydrogen society to cope with global environmental problems. JAERI has conducted R and D on HTGR as the world leader such as to achieve a reactor outlet coolant temperature of 950 degC in the HTTR (High Temperature Engineering Test Reactor) in April 2004 as the world's first and also to succeed in continuous hydrogen production with a bench-scale apparatus of closed cycle iodine-sulfur (IS) process for six and half hours in August 2003 as the world's first. Overview and present status of HTTR program were presented in details with background and main R and D results as well as international trend of HTGR development and future program on pilot tests facilities for hydrogen production demonstration in Japan. (T. Tanaka)

Construction of the HTTR and its testing program for advanced HTGR development

International Nuclear Information System (INIS)

Tanaka, T.; Baba, O.; Shiozawa, S.; Okubo, M.; Kunitomi, K.

1996-01-01

Concerning about global warming due to emission of greenhouse effect gas like CO 2 , it is essentially important to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of CO 2 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) which can supply high-temperature heat with high thermal efficiency as well as high heat-utilizing efficiency. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950 deg. C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). 2 refs, 2 figs, 1 tab., 2 photos

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

HTGR-GT closed-cycle gas turbine: a plant concept with inherent cogeneration (power plus heat production) capability

International Nuclear Information System (INIS)

McDonald, C.F.

1980-04-01

The high-grade sensible heat rejection characteristic of the high-temperature gas-cooled reactor-gas turbine (HTGR-GT) plant is ideally suited to cogeneration. Cogeneration in this nuclear closed-cycle plant could include (1) bottoming Rankine cycle, (2) hot water or process steam production, (3) desalination, and (4) urban and industrial district heating. This paper discusses the HTGR-GT plant thermodynamic cycles, design features, and potential applications for the cogeneration operation modes. This paper concludes that the HTGR-GT plant, which can potentially approach a 50% overall efficiency in a combined cycle mode, can significantly aid national energy goals, particularly resource conservation

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

Overview of HTGR heat utilization system development at JAERI

International Nuclear Information System (INIS)

Miyamoto, Y.; Shiozawa, S.; Ogawa, M.; Akino, N.; Shimizu, S.; Hada, K.; Inagaki, Y.; Onuki, K.; Takeda, T.; Nishihara, T.

1998-01-01

The Japan Atomic Energy Research Institute (JAERI) has conducted research and development of nuclear heat utilization systems of a High Temperature Gas cooled Reactor (HTGR), which are capable to meet a large amount of energy demand without significant CO 2 emission to relax the global warming issue. The High Temperature engineering Test Reactor (HTTR) with thermal output of 30 MW and outlet coolant temperature of 950 deg C, the first HTGR in Japan, is under construction on the JAERI site, and its first criticality is scheduled for mid-1998. After the reactor performance and safety demonstration tests for several years, a hydrogen production system will be connected to the HTTR. A demonstration program on hydrogen production started in January 1997, in JAERI, as a study consigned by the Science and Technology Agency. A hydrogen production system connected to the HTTR is designed to be able to produce hydrogen by steam reforming of natural gas, using nuclear heat of 10 MW from the HTTR. The safety principle and standard are investigated for the HTTR hydrogen production system. In order to confirm safety, controllability and performance of key components in the HTTR hydrogen production system, an out-of-pile test facility on the scale of approximately 1/30 of the HTTR hydrogen production system is installed. It is equipped with an electric heater as a heat source instead of the HTTR. The out-of-pile test will be performed for four years after 2001. The HTTR hydrogen production system will be demonstratively operated after 2005 at its earliest plan. Other basic studies on the hydrogen production system using thermochemical water splitting, an iodine sulphur (IS) process, and technology of distant heat transport with microencapsulated phase change material have been carried out for more effective and various uses of nuclear heat. (author)

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.

Design and thermal dynamic analyses on the intermediate heat exchanger for HTGR

International Nuclear Information System (INIS)

Mori, M.; Mizuno, M.; Ito, M.; Urabe, S.

1986-01-01

The intermediate heat exchanger (IHX), one of the most important components in the high temperature gas cooled reactor (HTGR), is a high performance helium/helium (He/He) heat exchanger operated at a very high temperature above 900 0 C to transmit the nuclear heat from the reactor core to the nuclear heat utilization systems such as the chemical plant. Having to meet, in addition, the requirement of the pressure boundary as the Class-1 it demands the accurate estimation of thermal performance and analytical prediction of thermal behaviors to secure its integrity throughout the service life. In the present works, the newly-developed analytical codes carry out designing thermal performance and analyzing dynamic thermal behaviors of the IHX. These codes have been developed on a great deal of data and studies related to the research and development on the 1.5 MWt- and the 25 MWt-IHXs. This paper shows the design on the IHX, the results of the dynamic analyses on the 1.5 MWt-IHX with the comparison to the experimental data and the analytical predictions of the dynamic thermal behaviors on the 25 MWt-IHX. The results calculated are in fairly good agreement with the experimental data on the 1.5 MWt-IHX, the fact that has verified the analytical codes to be reasonable and much useful for the thermal design of the IHX. These presented results and data are available for the design of the IHX of HTGR

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

SONATINA-2H: a computer program for seismic analysis of the two-dimensional horizontal slice HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi

1990-02-01

A Computer program SONATINA-2H has been developed for predicting the behavior of a two-dimensional horizontal HTGR core under seismic excitation. SONATINA-2H is a general two-dimensional computer program capable of analyzing the horizontal slice HTGR core with the fixed side reflector blocks and its restraint structures and the core support structure. In the analytical model, each block is treated as a rigid body and represent one column of the reactor core and is connected to the core support structure by means of column springs and viscous dampers. A single dashpot model is used for the collision process between adjacent blocks. The core support structure is represented by a single block. The computer program SONATINA-2H is capable of analyzing the core behavior for an excitation input applied simultaneously in two mutually perpendicular horizontal directions. In the present report are given, the theoretical formulation of the analytical model, an user's manual to describe the input and output format and sample problems. (author)

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

Evaluation of the significance of inverse oxidation for HTGR graphites

International Nuclear Information System (INIS)

Lee, B.S.; Heiser, J. III; Sastre, C.

1983-01-01

The inverse oxidation refers to a higher mass loss inside the graphite than the outside. In 1980, Wichner et al reported this phenomenon (referred to as inside/out corrosion) observed in some H451 graphites, and offered an explanation that a catalyst (almost certainly Fe) is activated by the progressively increasing reducing conditions found in the graphite interior. Recently, Morgan and Thomas (1982) investigated this phenomenon is PGX graphites, and agreed on the existing mechanism to explain this pheomenon. They also called for attention to the possibility that this phenomenon may occur under HTGR (High Temperature Gas-Cooled Reactor) operating conditions. The purpose of this paper is to confirm the above mentioned explanation for this phenomenon and to evaluate the significance of this effect for HTGR graphites under realistic reactor conditions

Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

International Nuclear Information System (INIS)

Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

1977-04-01

The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

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)

Recent evolution of HTGR instrumentation in the USA

International Nuclear Information System (INIS)

Rodriguez, C.

1982-06-01

The reactor instrumentation system for the 2240 MW(t) HTGR includes ex-core neutron detectors for automatic nuclear power control, separate ex-core neutron detectors for automatic protection purposes (reactor trip), reactor core outlet thermocouples that measure the temperature of the primary coolant (helium) as it exits the nuclear core, cold helium thermocouples that measure the temperature of the primary coolant as it enters the core, external pressure differential gages that measure primary coolant flow, in-core fission chambers that are utilized to map neutron flux, and ex-core primary coolant moisture monitors. All of these subsystems, except for the in-core flux mapping units, are also part of the Fort St. Vrain HTGR, which has provided significant experience for the design of the new system. In-core flux mapping is not necessary at FSV for normal operation because its relatively small core is fairly ''visible'' from the location of the ex-core instruments. However, temporary in-core fission couples, microphones, and displacement sensors, as well as sensitive ex-core accelerometers were utilized to identify periodic core block lateral movement and measure neutron flux and primary coolant temperatures. A search for in-core sensors to facilitate mapping neutron flux distributions in the larger core of the 2240 MW(t) HTGR has led to the selection of a high temperature fission chamber, which has been tested up to 1000 deg. C at General Atomic. The chamber shows adequate signal to noise ratio and repeatability. Other reactor instruments planned for the 2240 MW(t) are of the FSV type (i.e. thermocouples) or improved versions of the FSV design (i.e. moisture monitors). New concepts such as acoustic thermometers are also being considered

Sensitivity and Uncertainty Analysis of IAEA CRP HTGR Benchmark Using McCARD

International Nuclear Information System (INIS)

Jang, Sang Hoon; Shim, Hyung Jin

2016-01-01

The benchmark consists of 4 phases starting from the local standalone modeling (Phase I) to the safety calculation of coupled system with transient situation (Phase IV). As a preliminary study of UAM on HTGR, this paper covers the exercise 1 and 2 of Phase I which defines the unit cell and lattice geometry of MHTGR-350 (General Atomics). The objective of these exercises is to quantify the uncertainty of the multiplication factor induced by perturbing nuclear data as well as to analyze the specific features of HTGR such as double heterogeneity and self-shielding treatment. The uncertainty quantification of IAEA CRP HTGR UAM benchmarks were conducted using first-order AWP method in McCARD. Uncertainty of the multiplication factor was estimated only for the microscopic cross section perturbation. To reduce the computation time and memory shortage, recently implemented uncertainty analysis module in MC wielandt calculation was adjusted. The covariance data of cross section was generated by NJOY/ERRORR module with ENDF/B-VII.1. The numerical result was compared with evaluation result of DeCART/MUSAD code system developed by KAERI. IAEA CRP HTGR UAM benchmark problems were analyzed using McCARD. The numerical results were compared with Serpent for eigenvalue calculation and DeCART/MUSAD for S/U analysis. In eigenvalue calculation, inconsistencies were found in the result with ENDF/B-VII.1 cross section library and it was found to be the effect of thermal scattering data of graphite. As to S/U analysis, McCARD results matched well with DeCART/MUSAD, but showed some discrepancy in 238U capture regarding implicit uncertainty.

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

NGNP Component Test Capability Design Code of Record

Energy Technology Data Exchange (ETDEWEB)

S.L. Austad; D.S. Ferguson; L.E. Guillen; C.W. McKnight; P.J. Petersen

2009-09-01

The Next Generation Nuclear Plant Project is conducting a trade study to select a preferred approach for establishing a capability whereby NGNP technology development testing—through large-scale, integrated tests—can be performed for critical HTGR structures, systems, and components (SSCs). The mission of this capability includes enabling the validation of interfaces, interactions, and performance for critical systems and components prior to installation in the NGNP prototype.

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

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

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

Application of the lines of protection concept to the HTGR-SC/C

International Nuclear Information System (INIS)

1981-09-01

This study of the application of the line of protection (LOP) concept to high temperature gas-cooled reactors (HTGRs) was motivated by a desire to develop a simple and straightforward HTGR safety concept that embodies many of the more complicated and seemingly conflicting concepts facing nuclear industry safety today. These concepts include: (1) defense in depth; (2) design basis events; (3) core damage events (degraded cores); (4) probabilistic analysis and risk assessment; (5) numerical safety goals; and (6) plant investment protection. The LOP concept described herein attempts to incorporate many of the important principles of each into a cohesive framework which provides an overall logic, meaning, and direction for conducting HTGR design and research activities

Proceedings of the 1st JAERI symposium on HTGR technologies

International Nuclear Information System (INIS)

1990-07-01

This report was edited as the Proceedings of the 1st JAERI Symposium on HTGR Technologies, - Design, Licensing Requirements and Supporting Technologies -, collecting the 21 papers presented in the Symposium. The 19 of the presented papers are indexed individually. (J.P.N.)

60-MW/sub t/ methanation plant design for HTGR process heat

International Nuclear Information System (INIS)

Davis, C.R.; Arcilla, N.T.; Hui, M.M.; Hutchins, B.A.

1982-07-01

This report describes a 60 MW(t) Methanation Plant for generating steam for industrial applications. The plant consists of four 15 MW(t) methanation trains. Each train is connected to a pipeline and receives synthesis gas (syngas) from a High Temperature Gas-Cooled Reactor Reforming (HTGR-R) plant. Conversion of the syngas to methane and water releases exothermic heat which is used to generate steam. Syngas is received at the Methanation Plant at a temperature of 80 0 F and 900 psia. One adiabatic catalytic reactor and one isothermal catalytic reactor, in each methanation train, converts the syngas to 92.2% (dry bases) methane. Methane and condensate are returned at temperatures of 100 to 125 0 F and at pressures of 860 to 870 psia to the HTGR-R plant for the reproduction of syngas

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

INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

Directory of Open Access Journals (Sweden)

Sudarmono Sudarmono

2015-03-01

Full Text Available The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR. Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology  (PTRKN has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generation system, which has high energy efficiency, and has high and clean inherent safety level. The geometry and structure of the HTGR200 core are designed to produce the output of helium gas coolant temperature as high as 950 °C to be used for hydrogen production and other industrial processes in co-generative way. The output of very high temperature helium gas will cause thermal stress on the fuel pebble that threats the integrity of fission product confinement. Therefore, it is necessary to perform thermal-flow evaluation to determine the temperature distribution in the graphite and fuel pebble in the HTGR core. The evaluation was carried out by Thermix-Konvek module code that has been already integrated into VSOP'94 code. The HTGR core geometry was done using BIRGIT module code for 2-D model (RZ model with 5 channels of pebble flow in active core in the radial direction. The evaluation results showed that the highest and lowest temperatures in the reactor core are 999.3 °C and 886.5 °C, while the highest temperature of TRISO UO2 is 1510.20 °C in the position (z= 335.51 cm; r=0 cm. The analysis done based on reactor condition of 120 kg/s of coolant mass flow rate, 7 MPa of pressure and 200 MWth of power. Compared to the temperature distribution resulted between VSOP’94 code and fuel temperature limitation as high as 1600 oC, there is enough safety margin from melting or disintegrating. Keywords: Thermal-Flow, VSOP’94, Thermix-Konvek, HTGR, temperature

Availability of steam generator against thermal disturbance of hydrogen production system coupled to HTGR

International Nuclear Information System (INIS)

Shibata, Taiju; Nishihara, Tetsuo; Hada, Kazuhiko; Shiozawa, Shusaku

1996-01-01

One of the safety issues to couple a hydrogen production system to an HTGR is how the reactor coolability can be maintained against anticipated abnormal reduction of heat removal (thermal disturbance) of the hydrogen production system. Since such a thermal disturbance is thought to frequently occur, it is desired against the thermal disturbance to keep reactor coolability by means other than reactor scram. Also, it is thought that the development of a passive cooling system for such a thermal disturbance will be necessary from a public acceptance point of view in a future HTGR-hydrogen production system. We propose a SG as the passive cooling system which can keep the reactor coolability during a thermal disturbance of a hydrogen production system. This paper describes the proposed steam generator (SG) for the HTGR-hydrogen production system and a result of transient thermal-hydraulic analysis of the total system, showing availability of the SG against a thermal disturbance of the hydrogen production system in case of the HTTR-steam reforming hydrogen production system. (author)

Design codes for gas cooled reactor components

International Nuclear Information System (INIS)

1990-12-01

High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

Operational, control and protective system transient analyses of the closed-cycle GT-HTGR power plant

International Nuclear Information System (INIS)

Openshaw, F.L.; Chan, T.W.

1980-07-01

This paper presents a description of the analyses of the control/protective system preliminary designs for the gas turbine high-temperature gas-cooled reactor (GT-HTGR) power plant. The control system is designed to regulate reactor power, control electric load and turbine speed, control the temperature of the helium delivered to the turbines, and control thermal transients experienced by the system components. In addition, it provides the required control programming for startup, shutdown, load ramp, and other expected operations. The control system also handles conditions imposed on the system during upset and emergency conditions such as loop trip, reactor trip, or electrical load rejection

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.

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

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)

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)

Identification of key amino acid residues in the hTGR5-nomilin interaction and construction of its binding model.

Science.gov (United States)

Sasaki, Takashi; Mita, Moeko; Ikari, Naho; Kuboyama, Ayane; Hashimoto, Shuzo; Kaneko, Tatsuya; Ishiguro, Masaji; Shimizu, Makoto; Inoue, Jun; Sato, Ryuichiro

2017-01-01

TGR5, a member of the G protein-coupled receptor (GPCR) family, is activated by bile acids. Because TGR5 promotes energy expenditure and improves glucose homeostasis, it is recognized as a key target in treating metabolic diseases. We previously showed that nomilin, a citrus limonoid, activates TGR5 and confers anti-obesity and anti-hyperglycemic effects in mice. Information on the TGR5-nomilin interaction regarding molecular structure, however, has not been reported. In the present study, we found that human TGR5 (hTGR5) shows higher nomilin responsiveness than does mouse TGR5 (mTGR5). Using mouse-human chimeric TGR5, we also found that three amino acid residues (Q77ECL1, R80ECL1, and Y893.29) are important in the hTGR5-nomilin interaction. Based on these results, an hTGR5-nomilin binding model was constructed using in silico docking simulation, demonstrating that four hydrophilic hydrogen-bonding interactions occur between nomilin and hTGR5. The binding mode of hTGR5-nomilin is vastly different from those of other TGR5 agonists previously reported, suggesting that TGR5 forms various binding patterns depending on the type of agonist. Our study promotes a better understanding of the structure of TGR5, and it may be useful in developing and screening new TGR5 agonists.

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.

Research program of the high temperature engineering test reactor for upgrading the HTGR technology

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

1997-07-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

Granular effect on the effective cross sections in the HTGR type reactors

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de.

1975-01-01

Effective cross section of bars for HTGR is studied from the point of view of heterogeneity. Microscopical heterogeneity due to grains is represented by a self-shielding factor, which is well determined [pt

INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

OpenAIRE

Sudarmono Sudarmono

2015-01-01

The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR). Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology  (PTRKN) has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generati...

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)

A 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

This paper outlines a small nuclear plant concept which is not meant to replace the large nuclear power plants that will continue to be needed by the industrialized nations, but rather recognizes the needs of the smaller energy user, both for special applications in the US and for the developing nations. The small High-Temperature Gas-Cooled Reactor (HTGR), whose introduction will be very dependent on market forces, represents only one approach to meet these needs. The design of a small power plant that could be inherently safer and that might have costs less than those indicated by the traditional reverse-economy-of-scale effect is discussed. Topics considered include power plant economics, the small steam cycle HTGR thermodynamic cycle, the reactor nuclear heat source layout, the reactor heat removal system (main loop cooling, a vessel cooling system with reactor pressurized, vessel cooling system with reactor depressurized), safety considerations, investment risk protection, the technology base, and applications for the small HTGR plant concept

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

Preliminary risk assessments of the small HTGR

International Nuclear Information System (INIS)

Everline, C.J.; Bellis, E.A.

1985-05-01

Preliminary investment and safety risk assessments were performed for a preconceptual design of a four-module 250-MW(t) side-by-side steel-vessel pebble bed HTGR plant. Broad event spectra were analyzed involving component damage resulting in unscheduled plant outages and fission product releases resulting in offsite doses. The preliminary assessment indicates at this stage of the design that two categories of events govern the investment risk envelope: primary coolant leaks which release some circulating and plate-out activity that contaminates the confinement and turbogenerator damage which involves extensive turbine blade failure. Primary coolant leaks are important contributors because associated cleanup and decontamination requirements result in longer outages that arise from other events with comparable frequencies. Turbogenerator damage is the salient low-frequency investment risk accident due to the relatively long outages being experienced in the industry. Thermal transients are unimportant investment risk contributors because pressurized core heatups cause little damage, and depressurized core heatups occur at negligible frequencies relative to the forced outage goal. These preliminary results demonstrate investment and safety risk goal compliance at this stage in the design process. Studies are continuing in order to provide valuable insights into risk-significant events to assure a balanced approach to meeting user and regulatory requirements

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

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

Summary of ORNL work on NRC-sponsored HTGR safety research, July 1974-September 1980

International Nuclear Information System (INIS)

Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Delene, J.G.; Harrington, R.M.; Hatta, M.; Hedrick, R.A.; Johnson, L.G.; Sanders, J.P.

1982-03-01

A summary is presented of the major accomplishments of the Oak Ridge National Laboratory (ORNL) research program on High-Temperature Gas-Cooled Reactor (HTGR) safety. This report is intended to help the nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR. The ORNL program includes extensive work on dynamics-related safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification. Cooperative efforts with other programs are also described. Suggestions for near-term and long-term research are presented

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

International Nuclear Information System (INIS)

Baldwin, D.H.

1984-08-01

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

Scaling laws for HTGR core block seismic response

International Nuclear Information System (INIS)

Dove, R.C.

1977-01-01

This paper discusses the development of scaling laws, physical modeling, and seismic testing of a model designed to represent a High Temperature Gas-Cooled Reactor (HTGR) core consisting of graphite blocks. The establishment of the proper scale relationships for length, time, force, and other parameters is emphasized. Tests to select model materials and the appropriate scales are described. Preliminary results obtained from both model and prototype systems tested under simulated seismic vibration are presented

Time-dependent high-temperature low-cycle fatigue behavior of nickel-base heat-resistant alloys for HTGR

International Nuclear Information System (INIS)

Tsuji, Hirokazu; Kondo, Tatsuo

1988-06-01

A series of strain controlled low-cycle fatigue tests at 900 deg C in the simulated HTGR helium environment were conducted on Hastelloy X and its modified version, Hastelloy XR in order to examine time-dependent high-temperature low-cycle fatigue behavior. In the tests with the symmetric triangular strain waveform, decreasing the strain rate led to notable reductions in the fatigue life. In the tests with the trapezoidal strain waveform with different holding types, the fatigue life was found to be reduced most effectively in tensile hold-time experiments. Based on the observations of the crack morphology the strain holding in the compressive side was suggested to play the role of suppressing the initiation and the growth of internal cracks or cavities, and to cause crack branching. When the frequency modified fatigue life method and/or the prediction of life by use of the ductility were applied, both the data obtained with the symmetric triangular strain waveform and those with the tensile hold-time experiments lay on the straight line plots. The data, however, obtained with the compressive and/or both hold-time experiments could not be handled satisfactorily by those methods. When the cumulative damage rule was applied, it was found that the reliability of HTGR components was ensured by limiting the creep-fatigue damage fraction within the value of 1. (author)

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

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

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

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

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.

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.)

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

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.

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.

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

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)

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)

SONATINA-1: a computer program for seismic response analysis of column in HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi

1980-11-01

An computer program SONATINA-1 for predicting the behavior of a prismatic high-temperature gas-cooled reactor (HTGR) core under seismic excitation has been developed. In this analytical method, blocks are treated as rigid bodies and are constrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions. Coulomb friction between blocks and between dowel holes and pins is also considered. A spring dashpot model is used for the collision process between adjacent blocks and between blocks and boundary walls. Analytical results are compared with experimental results and are found to be in good agreement. The computer program can be used to predict the behavior of the HTGR core under seismic excitation. (author)

Safety and licensing analyses for the Fort St. Vrain HTGR

International Nuclear Information System (INIS)

Ball, S.J.; Conklin, J.C.; Harrington, R.M.; Cleveland, J.C.; Clapp, N.E. Jr.

1982-01-01

The Oak Ridge National Laboratory (ORNL) safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks during a postulated accident

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

Evaluation of creep-fatigue/ environment interaction in Ni-base wrought alloys for HTGR application

International Nuclear Information System (INIS)

Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira

1986-01-01

High Temperature Gas-cooled Reactor (HTGR) systems should be designed based on the high temperature structural strength design procedures. On the development of design code, the determination of failure criteria under cyclic loading and severe environments is one of the most important items. By using the previous experimental data for Ni-base wrought alloys, Inconel 617 and Hastelloy XR, several evaluation methods for creep-fatigue interaction were examined for their capability to predict their cyclic loading behavior for HTGR application. At first, the strainrange partitioning method, the frequency modified damage function and the linear damage summation rule were discussed. However, these methods were not satisfactory with the above experimental results. Thus, in this paper, a new fracture criterion, which is a modification of the linear damage summation rule, is proposed based on the experimental data. In this criterion, fracture is considered to occur when the sum of the fatigue damage, which is the function of the applied cyclic strain magnitude, and the modified creep damage, which is the function of the applied cyclic stress magnitude (determined as time devided by cyclic creep rupture time reflecting difference of creep damages by tensile creep and compressive creep), reaches a constant value. This criterion was successfully applied to the life prediction of materials at HTGR temperatures. (author)

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.

Development of THYDE-HTGR: computer code for transient thermal-hydraulics of high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Hirano, Masashi; Hada, Kazuhiko

1990-04-01

The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)

HTGR safety research at the Los Alamos National Laboratory

International Nuclear Information System (INIS)

Stroh, K.R.; Anderson, C.A.; Kirk, W.L.

1982-01-01

This paper summarizes activities undertaken at the Los Alamos National Laboratory as part of the High-Temperature Gas-Cooled Reactor (HTGR) Safety Research Program sponsored by the US Nuclear Regulatory Commission. Technical accomplishments and analysis capabilities in six broad-based task areas are described. These tasks are: fission-product technology, primary-coolant impurities, structural investigations, safety instrumentation and control systems, accident delineation, and phenomena modeling and systems analysis

Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

International Nuclear Information System (INIS)

Chow, J.G.Y.; Soo, P.; Epel, L.

1978-01-01

A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

Study of air ingress accident of an HTGR

International Nuclear Information System (INIS)

Hishida, Makoto

1995-01-01

Inherent properties of high temperature gas cooled reactors (HTGR) facilitate the design of HTGRs with high degree of passive safety performances. In this context, it is very important to establish a design criteria for a passive safe function for the air ingress accident. However, it is absolutely necessary to investigate the air ingress behavior during the accident before exploring the design criteria. The present paper briefly describes major activities and results of the air ingress research in our laboratory. (author)

A Benchmark Study of a Seismic Analysis Program for a Single Column of a HTGR Core

Energy Technology Data Exchange (ETDEWEB)

Kang, Ji Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

A seismic analysis program, SAPCOR (Seismic Analysis of Prismatic HTGR Core), was developed in Korea Atomic Energy Research Institute. The program is used for the evaluation of deformed shapes and forces on the graphite blocks which using point-mass rigid bodies with Kelvin-Voigt impact models. In the previous studies, the program was verified using theoretical solutions and benchmark problems. To validate the program for more complicated problems, a free vibration analysis of a single column of a HTGR core was selected and the calculation results of the SAPCOR and a commercial FEM code, Abaqus, were compared in this study.

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

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

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.)

Thermo-economic performance of HTGR Brayton power cycles

International Nuclear Information System (INIS)

Linares, J. L.; Herranz, L. E.; Moratilla, B. Y.; Fernandez-Perez, A.

2008-01-01

High temperature reached in High and Very High Temperature Reactors (VHTRs) results in thermal efficiencies substantially higher than those of actual nuclear power plants. A number of studies mainly driven by achieving optimum thermal performance have explored several layout. However, economic assessments of cycle power configurations for innovative systems, although necessarily uncertain at this time, may bring valuable information in relative terms concerning power cycle optimization. This paper investigates the thermal and economic performance direct Brayton cycles. Based on the available parameters and settings of different designs of HTGR power plants (GTHTR-300 and PBMR) and using the first and second laws of thermodynamics, the effects of compressor inter-cooling and of the compressor-turbine arrangement (i.e., single vs. multiple axes) on thermal efficiency have been estimated. The economic analysis has been based on the El-Sayed methodology and on the indirect derivation of the reactor capital investment. The results of the study suggest that a 1-axis inter-cooled power cycle has a similar thermal performance to the 3-axes one (around 50%) and, what's more, it is substantially less taxed. A sensitivity study allowed assessing the potential impact of optimizing several variables on cycle performance. Further than that, the cycle components costs have been estimated and compared. (authors)

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)

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).

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)

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

Energy Technology Data Exchange (ETDEWEB)

Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

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

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

Consideration on developing of leaked inflammable gas detection system for HTGR hydrogen production system

International Nuclear Information System (INIS)

Nishihara, Tetsuo; Nakamura, Masashi

1999-09-01

One of most important safety design issues for High Temperature Gas-cooled Reactor (HTGR) - Hydrogen Production System (HTGR-HPS) is to ensure reactor safety against fire and explosion at the hydrogen production plant. The inflammable gas mixture in the HTGR-HPS does not use oxygen in any condition and are kept in high pressure in the normal operation. The piping system and/or heat transfer tubes which have the potential possibility of combustible materials ingress into the Reactor Building (R/B) due to the failure are designed to prevent the failure against any events. Then, it is not necessary to consider their self-combustion in vessels nor leakage in the R/B. The only one case which we must consider is the ex-building fire or explosion caused by their leakage from piping or vessel. And it is important to mitigate their effects by means of early detection of gas leakage. We investigated our domestic standards on gas detection, applications of gas detectors, their detection principles, performance, sensitivity, reliability, their technical trends, and so on. We proposed three gas detection systems which may be applied in HTGR-HPS. The first one is the universal solid sensor system; it may be applied when there is no necessity to request their safety credits. The second is the combination of the improved solid sensor system and enhanced beam detector system; it may be applied when it is necessary to request their safety credit. And the third is the combination of the universal solid sensor system and the existing beam detector system; it may be applied when the plant owner request higher detector sensitivity than usual, from the view point of public acceptance, though there is not necessity to request their safety credits. To reduce the plant cost by refusing of safety credits to the gas leakage detection system, we proposed that the equipment required to isolate from others should be installed in the inertrized compartments. (author)

Conceptual design of small-sized HTGR system (4). Plant design and technical feasibility

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing L.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio

2013-09-01

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 2020s. HTR50S was designed for steam supply and electricity generation by the steam turbine with the reactor outlet temperature of 750degC as a reference plant configuration. On the other hand, the intermediate heat exchanger (IHX) will be installed in the primary loop to demonstrate the electricity generation by the helium gas turbine and hydrogen production by thermochemical water splitting by utilizing the secondary helium loop with the reactor outlet temperature of 900degC as a future plant configuration. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750degC to 900degC and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems (i.e., reactor internal components, reactor pressure vessel, vessel cooling system, shutdown cooling system, steam generator (SG), gas circulator, SG isolation and drainage system, reactor containment vessel, steam turbine and heat supply system) satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S. (author)

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

Ways to increase efficiency of the HTGR coupled with the gas-turbine power conversion unit - HTR2008-58274

International Nuclear Information System (INIS)

Golovko, V. F.; Kodochigov, N. G.; Vasyaev, A. V.; Shenoy, A.; Baxi, C. B.

2008-01-01

reactor plants with highly recuperative steam cycle with supercritical heat parameters, the net efficiency of electricity generation reaches 50-55%. There are three methods of Brayton cycle carnotization: regeneration, helium cooldown during compression, and heat supply during expansion. These methods can be used both separately and in combination, which gives a total of seven complex heat flow diagrams. Besides, there are ways to increase helium temperature at the reactor inlet and outlet, to reduce hydraulic losses in the helium path, to increase the turbomachine (TM) rotation speed in order to improve the turbine and compressor efficiency, to reduce helium leaks in the circulation path, etc. The analysis of GT-MHR, PBMR and GTHTR-300 development experience allows identification of the main ways of increasing the efficiency by selecting optimal parameters and design solutions for the reactor and power conversion unit. The paper estimates the probability of reaching the maximum electricity generation efficiency in reactor plants with the HTGR and gas turbine cycle with account of the up-to-date development status of major reactor plant components (reactor, vessels, turbo-compressor (TC), generator, heat exchange equipment, and structural materials). (authors)

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.

Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

International Nuclear Information System (INIS)

Demick, L.E.

2011-01-01

This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

Energy Technology Data Exchange (ETDEWEB)

L.E. Demick

2011-10-01

This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

Studies of iodine adsorption and desorption on HTGR coolant circuit materials

International Nuclear Information System (INIS)

Osborne, M.F.; Compere, E.L.; de Nordwall, H.J.

1976-04-01

Safety studies of the HTGR system indicate that radioactive iodine, released from the fuel to the helium coolant, may pose a problem of concern if no attenuation of the amount of iodine released occurs in the coolant circuit. Since information on iodine behavior in this system was incomplete, iodine adsorption on HTGR materials was studied in vacuum as a function of iodine pressure and of adsorber temperature. Iodine coverages on Fe 3 O 4 and Cr 2 O 3 approached maxima of about 2 x 10 14 and 1 x 10 14 atoms/cm 2 , respectively, whereas the iodine coverage on graphite under similar conditions was found to be less by a factor of about 100. Iodine desorption from the same materials into vacuum or flowing helium was investigated, on a limited basis, as a function of iodine coverage, of adsorber temperature, and of dry vs wet helium. The rate of vacuum desorption from Fe 3 O 4 was related to the spectrum of energies of the adsorption sites. A small amount of water vapor in the helium enhanced desorption from iron powder but appeared to have less effect on desorption from the metal oxides

Project summary plan for HTGR recycle reference facility

International Nuclear Information System (INIS)

Baxter, B.J.

1979-11-01

A summary plan is introduced for completing conceptual definition of an HTGR Recycle Reference Facility (HRRF). The plan describes a generic project management concept, often referred to as the requirements approach to systems engineering. The plan begins with reference flow sheets and provides for the progressive evolution of HRRF requirements and definition through feasibility, preconceptual, and conceptual phases. The plan lays end-to-end all the important activities and elements to be treated during each phase of design. Identified activities and elements are further supported by technical guideline documents, which describe methodology, needed terminology, and where relevant a worked example

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.)

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

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.)

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

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

Safety concerns and suggested design approaches to the HTGR Reformer process concept

Energy Technology Data Exchange (ETDEWEB)

Green, R.C.

1981-09-01

This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept.

Safety concerns and suggested design approaches to the HTGR Reformer process concept

International Nuclear Information System (INIS)

Green, R.C.

1981-09-01

This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept

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

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

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

The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

Energy Technology Data Exchange (ETDEWEB)

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

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.

Thermal cycle efficiency of the indirect combined HTGR-GT power generation system

Energy Technology Data Exchange (ETDEWEB)

Muto, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-02-01

High thermal efficiency of 50% could be expected in a power generation system coupling a high temperature gas-cooled reactor(HTGR) with a closed cycle gas turbine(GT). There are three candidate systems such as a direct cycle(DC), an indirect cycle(ICD) and an indirect combined cycle(IDCC). The IDCC could solve many problems in both the DC and the IDC and consists of a primary circuit and a secondary circuit where a topping cycle is a Brayton cycle and a bottoming cycle is a steam cycle. In this report, the thermal cycle efficiency of the IDCC is examined regarding configurations of components and steam pressure. It has been shown that there are two types of configurations, that is, a perfect cascade type and a semi-cascade one and the latter can be further classified into Case A, Case B and Case C. The conditions achieving the maximum thermal cycle efficiency were revealed for these cases. In addition, the optimum system configurations were proposed considering the thermal cycle efficiency, safety and plant arrangement. (author).

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)

Present activity of the feasibility study of HTGR-GT system

International Nuclear Information System (INIS)

Muto, Y.; Miyamoto, Y.; Shiozawa, S.

2001-01-01

In JAERI a feasibility study of the High Temperature Gas-cooled Reactor-Gas Turbine (HTGR-GT) system has been carried out since January, 1997 as an assigned work by the Science and Technology Agency. The study aims at obtaining a promising concept of HTGR-GT system that yields a high thermal efficiency and at the same time is economically competitive. Designs of a few candidate systems will be undertaken and their power generation costs will be evaluated in parallel with design works, some experimental works such as the fabrication of a plate-fin type heat exchanger core and material tests will be carried out. The study will be continued till 2000 fiscal year. In 1997 fiscal year, a preliminary design of a direct cycle plant of 600 MWt was developed. A reactor inlet gas temperature of 460 deg. C, a reactor outlet gas temperature of 850 deg. C and a helium gas pressure of 6MPa were selected. Some advanced technologies were adopted such as a monolithic fuel compact and a control rod sheath made of carbon/carbon composite material. They were very effective to enhance the heat transfer of fuel and to reduce the core bypass flow. As a result, a power density of 6MW/m 3 and the maximum burnup of 10 5 MWD/ton were achieved. A single-shaft horizontal turbomachine of 3600 rpm was selected to ease the mechanical design of the rotor supported by magnetic bearings. The turbine, two compressors, a generator and six units of intercooler were placed in a turbine vessel, Plate-fin type recuperator and precooler are installed in a vertical heat exchanger vessel. By this design, a net thermal efficiency of 45.7% is expected to be achieved. To develop a high performance plate-fin recuperator, a core model of W200 mm x L200 mm x H200 mm with small fin size of 1.15 mm height was fabricated and as a result of tests, leak tightness, component strength and bonding appearance were found to be satisfactory. In 1998 fiscal year, a design of a direct cycle plant of 300 MWt is undertaken. The

Passive afterheat removal in the HTGR with the liner cooling system as a heat sink

International Nuclear Information System (INIS)

Rehm, W.; Jahn, W.; Verfondern, K.

1984-09-01

The report deals with the transients of temperature and system pressure and the fission product behaviour in the primary circuit of an HTGR during passive afterheat removal, where the liner cooling system of the PCRV serves as a heat sink. The analysis has been made for the PNP-500-reactor representing nuclear plants with medium thermal power. The investigations show that the liner cooling system is able to control a core heatup. High temperature loads are encountered in the upper core region. In the case of a reactor under pressure the fuel elements and the primary circuit remain intact as the first and second barriers for fission products. In the case of a depressurized primary circuit the liner cooling system also keeps the PCRV at normal operating temperatures. The effects of a core heatup on component damage and release of fission products are thus limited. (orig.) [de

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)

Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

International Nuclear Information System (INIS)

Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

2014-01-01

Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

Progress of independent feasibility study for modular HTGR demonstration plant to be built in China

International Nuclear Information System (INIS)

He Jiachen

1989-01-01

Many regions in China are suffering from shortage of energy as a result of the rapid growth of the national economy, for example, the growth rate of national production in 1988 reached 11.2%. A great number of coal fired plants have been built in many industrial areas. However, the difficulties relating to the transportation of coal and environmental pollution have become more and more serious. The construction of hydropower plants is limited due to uneven geographic conditions and seasons. For these reasons China needs to develop nuclear power plants. Nowadays, it has been decided, that PWR will be the main reactor type in our country, but in some districts or under some conditions modular HTGR may have distinct advantages and become an attractive option. The possible plant site description and preliminary result of economic analysis of modular HTGR type reactor are briefly discussed in this presentation

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

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

Preliminary Study on the Development of MIDAS/GCR to Simulate the Plate-out Phenomena from a HTGR

International Nuclear Information System (INIS)

Park, Jong-Hwa; Kim, Dong-Ha; Lee, Won-Jae

2006-01-01

In HTGR, the dominant removal mechanism of the condensable fission product gas is a 'plate-out' on various kinds of surfaces over the primary coolant loop. The plate-outs are complex phenomena that are dependent on the mass transfer rate from the coolant to the fixed surface, the adsorption and desorption of the gas fission product, the material of the surfaces, the operation temperature, the fission product species, etc. In a normal operation, the important information on a plate-out is the amount and the distribution and the type of isotope. This information is applied to construct a safety engineering system, to calculate the necessary shielding and to estimate the impact on the environment. The status of a model development and available data are performed extensively but the data still has a large uncertainty. The objective of this study is to compare the condensation model of a gas fission product in the MIDAS for a PWR with the PADLOC model for a HTGR developed by GA and to perform a feasibility calculation on OGL-1 with MIDAS. The results of the model review on MIDAS and PADLOC, the feasibility calculation results on OGL-1 with MIDAS and the phenomena to be implemented into MIDAS to simulate the plate-out phenomena from HTGR are identified and listed

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).

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.

Role of the HTGR in the U.S. industrial energy market

International Nuclear Information System (INIS)

Leeth, G.G.

1981-01-01

The HTGR is considered for a variety of applications to the U.S. industrial energy markets. These include a number of synfuel processes, shale oil conversion, methanol production, ammonia production, and both open and closed-loop pipeline systems. Potential market size appears to be approximately 300-400 GW (t) in the 2000 to 2020 time period. In addition to potential cost advantages, the closed-loop nuclear system has several significant advantages over alternative fossil systems. 5 refs

The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

International Nuclear Information System (INIS)

Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

1997-01-01

Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

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)

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)

Design of performance and analysis of dynamic and transient thermal behaviors on the intermediate heat exchanger for HTGR

International Nuclear Information System (INIS)

Mori, Michitsugu; Mizuno, Minoru; Itoh, Mitsuyoshi; Urabe, Shigemi

1985-01-01

The intermediate heat exchanger (IHX) is designed as the high temperature heat exchanger for HTGR (High Temperature Gas-cooled Reactor), which transmits the primary coolant helium's heat raised up to about 950 0 C in the reactor core to the secondary helium or the nuclear heat utilization. Having to meet, in addition, the requirement of the primary coolant pressure boundary as the Class-1 component, it must be secured integrity throughout the service life. This paper will show (1) the design of the thermal performance; (2) the results of the dynamic analyses of the 1.5 MWt-IHX with its comparison to the experimental data; (3) the analytical predictions of the dynamic thermal behaviors under start-up and of the transient thermal behaviors during the accident on the 25 MWt-IHX. (author)

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

Nondestructive evaluation of the oxidation and strength of the Fort Saint Vrain HTGR support block

International Nuclear Information System (INIS)

Tingey, G.L.; Posakony, G.J.; Morgan, W.C.; Prince, J.M.; Hill, R.W.; Lessor, D.L.

1982-04-01

Non-destructive detection of changes in the strength of graphite support structures in a HTGR appears to be feasible using sonic velocity measurements where access for through transmission is possible. Therefore, future HTGR designs should consider providing such access. Where access is not available, strength changes can be correlated with oxidation profiles in the support member. These oxidation profiles can be determined non-destructively by a combination of eddy current measurements to detect near surface oxidation and sonic backscattering measurements designed to determine oxidation in depth. The Fort Saint Vrain reactor provides an operating reactor to test the applicability of the eddy current and sonic backscattering techniques for determination of oxidation in a support block. Furthermore, such tests in Fort Saint Vrain will supply base line data which will be useful in assuring an adequate strength of the support structure for the lifetime of the reactor. Equipment is, therefore, being developed for tests to be conducted during the next major refueling of the reactor

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)

Assessment of modelling needs for safety analysis of current HTGR concepts

International Nuclear Information System (INIS)

Kroeger, P.G.; Van Tuyle, G.J.

1985-12-01

In view of the recent shift in emphasis of the DOE/Industry HTGR development efforts to smaller modular designs it became necessary to review the modelling needs and the codes available to assess the safety performance of these new designs. This report provides a final assessment of the most urgent modelling needs, comparing these to the tools available, and outlining the most significant areas where further modelling is required. Plans to implement the required work are presented. 47 refs., 20 figs

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.)

Three-dimensional computer code for the nonlinear dynamic response of an HTGR core

International Nuclear Information System (INIS)

Subudhi, M.; Lasker, L.; Koplik, B.; Curreri, J.; Goradia, H.

1979-01-01

A three-dimensional dynamic code has been developed to determine the nonlinear response of an HTGR core. The HTGR core consists of several thousands of hexagonal core blocks. These are arranged inlayers stacked together. Each layer contains many core blocks surrounded on their outer periphery by reflector blocks. The entire assembly is contained within a prestressed concrete reactor vessel. Gaps exist between adjacent blocks in any horizontal plane. Each core block in a given layer is connected to the blocks directly above and below it via three dowell pins. The present analystical study is directed towards an invesstigation of the nonlinear response of the reactor core blocks in the event of a seismic occurrence. The computer code is developed for a specific mathemtical model which represents a vertical arrangement of layers of blocks. This comprises a block module of core elements which would be obtained by cutting a cylindrical portion consisting of seven fuel blocks per layer. It is anticipated that a number of such modules properly arranged could represent the entire core. Hence, the predicted response of this module would exhibit the response characteristics of the core

Boat sampling technique for assessment of ageing of components

International Nuclear Information System (INIS)

Kumar, Kundan; Shyam, T.V.; Kayal, J.N.; Rupani, B.B.

2006-01-01

Boat sampling technique (BST) is a surface sampling technique, which has been developed for obtaining, in-situ, metal samples from the surface of an operating component without affecting its operating service life. The BST is non-destructive in nature and the sample is obtained without plastic deformation or without thermal degradation of the parent material. The shape and size of the sample depends upon the shape of the cutter and the surface geometry of the parent material. Miniature test specimens are generated from the sample and the specimens are subjected to various tests, viz. Metallurgical Evaluation, Metallographic Evaluation, Micro-hardness Evaluation, sensitisation test, small punch test etc. to confirm the integrity and assessment of safe operating life of the component. This paper highlights design objective of boat sampling technique, description of sampling module, sampling cutter and its performance evaluation, cutting process, boat samples, operational sequence of sampling module, qualification of sampling module, qualification of sampling technique, qualification of scooped region of the parent material, sample retrieval system, inspection, testing and examination to be carried out on the boat samples and scooped region. (author)

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

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

Application of modern control theory to HTGR-plant

International Nuclear Information System (INIS)

Izaki, Makoto; Kubo, Hiroaki; Yamazaki, Eiji; Suzuki, Katsuo.

1988-01-01

The classical control theory approach to the multivariate control problem is to decouple the system intentionally and to treat each loop independently. As a result, final control system design is limited in complexity by the available mathematical techniques limitation and it's control performance is insufficient in many cases. The modern control theory approach based on the state variables to the problem provides far more powerful methods and more design flexibility than the classical control theory approach by the new mathematical formulation about the problem. The state variable feedback in formulating as an optimal regulator is the most effective way to obtain the desired control performance. In this report, some results of optimal regulator application to High Temperature Gas Cooled Reactor (HTGR) are shown. (author)

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

Treatment of operator actions in the HTGR risk assessment study

International Nuclear Information System (INIS)

Fleming, K.N.; Silady, F.A.; Hannaman, G.W.

1979-12-01

Methods are presented for the treatment of operator actions, developed in the AIPA risk assessment study. Some examples are given of how these methods were applied to the analysis of potential HTGR accidents. Realistic predictions of accident risks required a balanced treatment of both beneficial and detrimental actions and responses of human operators and maintenance crews. Th essential elements of the human factors methodology used in the AIPA study include event tree and fault tree analysis, time-dependent operator response and repair models, a method for quantifying common cause failure probabilities, and synthesis of relevant experience data for use in these models

TRANTHAC-1: transient thermal-hydraulic analysis code for HTGR core of multi-channel model

International Nuclear Information System (INIS)

Sato, Sadao; Miyamoto, Yoshiaki

1980-08-01

The computer program TRANTHAC-1 is for predicting thermal-hydraulic transient behavior in HTGR's core of pin-in-block type fuel elements, taking into consideration of the core flow distribution. The program treats a multi-channel model, each single channel representing the respective column composed of fuel elements. The fuel columns are grouped in flow control regions; each region is provided with an orifice assembly. In the region, all channels are of the same shape except one channel. Core heat is removed by downward flow of the control through the channel. In any transients, for given time-dependent power, total core flow, inlet coolant temperature and coolant pressure, the thermal response of the core can be determined. In the respective channels, the heat conduction in radial and axial direction are represented. And the temperature distribution in each channel with the components is calculated. The model and usage of the program are described. The program is written in FORTRAN-IV for computer FACOM 230-75 and it is composed of about 4,000 cards. The required core memory is about 75 kilowords. (author)

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

International Nuclear Information System (INIS)

1980-05-01

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

Assessment of the SRI Gasification Process for Syngas Generation with HTGR Integration -- White Paper

Energy Technology Data Exchange (ETDEWEB)

A.M. Gandrik

2012-04-01

This white paper is intended to compare the technical and economic feasibility of syngas generation using the SRI gasification process coupled to several high-temperature gas-cooled reactors (HTGRs) with more traditional HTGR-integrated syngas generation techniques, including: (1) Gasification with high-temperature steam electrolysis (HTSE); (2) Steam methane reforming (SMR); and (3) Gasification with SMR with and without CO2 sequestration.

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

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

A three-dimensional computer code for the nonlinear dynamic response of an HTGR core

International Nuclear Information System (INIS)

Subudhi, M.; Lasker, L.; Koplik, B.; Curreri, J.; Goradia, H.

1979-01-01

A three-dimensional dynamic code has been developed to determine the nonlinear response of an HTGR core. The HTGR core consists of several thousands of hexagonal core blocks. These are arranged in layers stacked together. Each layer contains many core blocks surrounded on their outer periphery by reflector blocks. The entire assembly is contained within a prestressed concrete reactor vessel. Gaps exist between adjacent blocks in any horizontal plane. Each core block in a given layer is connected to the blocks directly above and below it via three dowell pins. The present analytical study is directed towards an investigation of the nonlinear response of the reactor core blocks in the event of a seismic occurrence. The computer code is developed for a specific mathematical model which represents a vertical arrangement of layers of blocks. This comprises a 'block module' of core elements which would be obtained by cutting a cylindrical portion consisting of seven fuel blocks per layer. It is anticipated that a number of such modules properly arranged could represent the entire core. Hence, the predicted response of this module would exhibit the response characteristics of the core. (orig.)

Pre elementary design of primary reformer for hydrogen plant coupled with HTGR type NPP

International Nuclear Information System (INIS)

Dedy Priambodo; Erlan Dewita; Sudi Ariyanto

2012-01-01

Hydrogen has a high potent for new energy, because of it availability. Steam reforming is a fully developed commercial technology and is the most economical method for production of hydrogen. Steam reforming uses an external source of hot gas to heat tubes in which a catalytic reaction takes place that converts steam and lighter hydrocarbons such as natural gas (methane) or refinery feedstock into hydrogen and carbon monoxide (syngas) at high temperature on primary reformer (800-900°C). Utilization of helium from HTGR as heating medium for primary reformer has consequence to type and shape of its reactor. The main goal of this paper is to determine type/shape and pre elementary design of chemical reactor for the cogeneration system of Hydrogen Plant and HTGR The primary reformer for this system is Fixed Bed Multitube reactor with specification tube: NPS 3,5 Sch 40 ST 40S, 0.281 in thickness, number of tube 849 pieces and ASTM HH 30 for tube material. Tube arrangement is 'triangular pitch' on shell Split-Ring Floating Head from Steel Alloy SA 301 Grade B equipted with 8 baffles. (author)

Life time test of a partial model of HTGR helium-helium heat exchanger

International Nuclear Information System (INIS)

Kitagawa, Masaki; Hattori, Hiroshi; Ohtomo, Akira; Teramae, Tetsuo; Hamanaka, Junichi; Itoh, Mitsuyoshi; Urabe, Shigemi

1984-01-01

Authors had proposed a design guide for the HTGR components and applied it to the design and construction of the 1.5 Mwt helium heat exchanger test loop for the nuclear steel making under the financial support of the Japanese Ministry of International Trade and Industry. In order to assure that the design method covers all the conceivable failure mode and has enough safety margin, a series of life time tests of partial model may be needed. For this project, three types of model tests were performed. A life time test of a partial model of the center manifold pipe and eight heat exchanger tubes were described in this report. A damage criterion with a set of material constants and a simplified method for stress-strain analysis for stub tube under three dimensional load were newly developed and used to predict the lives of each tube. The predicted lives were compared with the experimental lives and good agreement was found between the two. The life time test model was evaluated according to the proposed design guide and it was found that the guide has a safety factor of approximately 200 in life for this particular model. (author)

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

HTGR high temperature process heat design and cost status report. Volume II. Appendices

International Nuclear Information System (INIS)

1981-12-01

Information is presented concerning the 850 0 C IDC reactor vessel; primary cooling system; secondary helium system; steam generator; heat cycle evaluations for the 850 0 C IDC plant; 950 0 C DC reactor vessel; 950 0 C DC steam generator; direct and indirect cycle reformers; methanation plant; thermochemical pipeline; methodology for screening candidate synfuel processes; ECCG process; project technical requirements; process gas explosion assessment; HTGR program economic guidelines; and vendor respones

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

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

Study on system layout and component design in the HTTR hydrogen production system. Contract research

Energy Technology Data Exchange (ETDEWEB)

Nishihara, Tetsuo; Shimizu, Akira [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tanihira, Masanori [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Uchida, Shoji [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

2003-01-01

The global warming becomes a significant issue in the world so that it needs to reduce the CO{sub 2} emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO{sub 2} emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

Study on system layout and component design in the HTTR hydrogen production system. Contract research

International Nuclear Information System (INIS)

Nishihara, Tetsuo; Shimizu, Akira; Uchida, Shoji

2003-01-01

The global warming becomes a significant issue in the world so that it needs to reduce the CO 2 emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO 2 emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

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

HTGR high temperature process heat design and cost status report. Volume II. Appendices

Energy Technology Data Exchange (ETDEWEB)

None

1981-12-01

Information is presented concerning the 850/sup 0/C IDC reactor vessel; primary cooling system; secondary helium system; steam generator; heat cycle evaluations for the 850/sup 0/C IDC plant; 950/sup 0/C DC reactor vessel; 950/sup 0/C DC steam generator; direct and indirect cycle reformers; methanation plant; thermochemical pipeline; methodology for screening candidate synfuel processes; ECCG process; project technical requirements; process gas explosion assessment; HTGR program economic guidelines; and vendor respones.

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.)

Derivation of criteria for primary circuit activity in an HTGR

International Nuclear Information System (INIS)

Su, S.D.; Barsell, A.W.

1980-11-01

This paper derives specific criteria for the circulating and plateout activity in the primary circuit for a 2170-MW(t) high temperature gas-cooled reactor-gas turbine (HTGR-GT) plant. Results show that for a design basis, (1) the circulating activity should be limited to 14,000 Ci Kr-88 (a principal nuclide) to meet both offsite dose and containment access constraint during normal operation and depressurization accidents, and (2) the plateout inventories for those important nuclides affecting shutdown maintenance should not exceed 10,000 Ci Ag-110m, 45,000 Ci Cs-134 and 130,000 Ci Cs-137. This paper presents bases and methodology for deriving such criteria and compares them with light water reactors. 5 tables

Standard Practices for Sampling for Particles in Aerospace Fluids and Components

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 These practices cover sampling procedures for use in determining the particle cleanliness of liquids and liquid samples from components. Three practices, A, B, and C, have been developed on the basis of component geometry in order to encompass the wide variety of configurations. These practices establish guidelines to be used in preparing detailed procedures for sampling specific components. Note 1—The term cleanliness used in these practices refers to solid particles in the liquid. It does not generally cover other foreign matter such as gases, liquids, and products of chemical degradation. Cleanliness with respect to particulate contamination does not necessarily give any indication of the other types of contamination. 1.2 All components, regardless of application, may be tested provided (1) the fluid medium selected is completely compatible with the materials, packing and fluid used in the test component, and test apparatus, and (2) the fluid is handled in accordance with the manufacturer's recom...

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

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

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

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.)

Experimental determination of the K_oo 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.

Review of fatigue criteria development for HTGR core supports

International Nuclear Information System (INIS)

Ho, F.H.; Vollman, R.E.

1979-10-01

Fatigue criteria for HTGR core support graphite structure are presented. The criteria takes into consideration the brittle nature of the material, and emphasizes the probabilistic approach in the treatment of strength data. The stress analysis is still deterministic. The conventional cumulative damage approach is adopted here. A specified minimum S-N curve is defined as the curve with 99% probability of survival at a 95% confidence level to accommodate random variability of the material strength. A constant life diagram is constructed to reconcile the effect of mean stress. The linear damage rule is assumed to account for the effect of random cycles. An additional factor of safety of three on cycles is recommended. The uniaxial S-N curve is modified in the medium-to-high cycle range (> 2 x 10 3 cycles) for mutiaxial fatigue effects

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

Significance and prospects of the energo-technological usage of HTGR for nuclear power development in the beginning of the XXI century

International Nuclear Information System (INIS)

Breger, A.Kh.; Putilov, A.V.; Bogoyavlensky, R.G.; Glebov, V.P.

1993-01-01

This report describes the economic efficiency of atomic stations (HTGR plants). The realization of the complex (energy radiation-technological) using of nuclear fuel leads to the economically effective mastering of nuclear energy sources instead of organic ones for supplying inductry and municipal economy. It is necessary to include in the power engineering development programme, under the circumstances of fulfillment of requirements of safety and reliability, research and development by the end of the century of the pattern of complex unit on the base of the HTGR with spherical fuel elements and (by 2010-15), mastering the energy-technological plants in high-energy branches of industry and municipal economy. Solving the mentioned problems will make a perceptible contribution into scientific progress, will allow to fulfill the conversion of war industry, attract highly qualified specialists to solving the tasks of national economy

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

Overall simulation of a HTGR plant with the gas adapted MANTA code

International Nuclear Information System (INIS)

Emmanuel Jouet; Dominique Petit; Robert Martin

2005-01-01

Full text of publication follows: AREVA's subsidiary Framatome ANP is developing a Very High Temperature Reactor nuclear heat source that can be used for electricity generation as well as cogeneration including hydrogen production. The selected product has an indirect cycle architecture which is easily adapted to all possible uses of the nuclear heat source. The coupling to the applications is implemented through an Intermediate Heat exchanger. The system code chosen to calculate the steady-state and transient behaviour of the plant is based on the MANTA code. The flexible and modular MANTA code that is originally a system code for all non LOCA PWR plant transients, has been the subject of new developments to simulate all the forced convection transients of a nuclear plant with a gas cooled High Temperature Reactor including specific core thermal hydraulics and neutronics modelizations, gas and water steam turbomachinery and control structure. The gas adapted MANTA code version is now able to model a total HTGR plant with a direct Brayton cycle as well as indirect cycles. To validate these new developments, a modelization with the MANTA code of a real plant with direct Brayton cycle has been performed and steady-states and transients compared with recorded thermal hydraulic measures. Finally a comparison with the RELAP5 code has been done regarding transient calculations of the AREVA indirect cycle HTR project plant. Moreover to improve the user-friendliness in order to use MANTA as a systems conception, optimization design tool as well as a plant simulation tool, a Man- Machine-Interface is available. Acronyms: MANTA Modular Advanced Neutronic and Thermal hydraulic Analysis; HTGR High Temperature Gas-Cooled Reactor. (authors)

Multiregional coupled conduction--convection model for heat transfer in an HTGR core

International Nuclear Information System (INIS)

Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

1978-01-01

HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

Approach to the HTGR core outlet temperature measurements in the United States

International Nuclear Information System (INIS)

Franklin, R.; Rodriguez, C.

1982-06-01

The High Temperature Gas-Cooled Reactor (HTGR) constructed at Fort St. Vrain Colorado (330 MWe) used Geminol thermocouples to measure the primary coolant temperature at the core outlet. The primary coolant (helium) is heated by the graphite core to temperatures in the range of 700 deg. to 750 deg. C. The combination of the high temperature, high flow rate and radiation at the core outlet area makes it difficult to obtain accurate temperature measurements. The Geminol thermocouples installed in the Fort St. Vrain reactor have provided accurate data for several years of power operation without any failures. The indicated temperature of the core outlet thermocouples agrees with a ''traversing'' thermocouple measurement to within +-2 deg. C. The Geminol thermocouple wire was provided by the Driver-Harris Company and is similar to the chromel versus alumel thermocouple. Geminol wire is no longer distributed and on future designs, chromel versus alumel wire will be used. The next large HTGR design, which is being performed with funding support from the United States Department of Energy, will incorporate replaceable thermocouples. The thermocouples used in the Fort St. Vrain reactor were permanently installed and large in diameter (6.35 mm) to insure good reliability. The replaceable thermocouples to be used in the next large reactor will be smaller in diameter (3.18 mm). These replaceable thermocouples will be inserted into the core outlet area through long curved guide tubes that are permanently installed. These guide tubes are as long as 18 meters and must be curved to reach the core outlet regions. Tests were conducted to prove that the thermocouples could be inserted and removed through the long curved guide tubes. (author)

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)

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

Sampling methods to the statistical control of the production of blood components.

Science.gov (United States)

Pereira, Paulo; Seghatchian, Jerard; Caldeira, Beatriz; Santos, Paula; Castro, Rosa; Fernandes, Teresa; Xavier, Sandra; de Sousa, Gracinda; de Almeida E Sousa, João Paulo

2017-12-01

The control of blood components specifications is a requirement generalized in Europe by the European Commission Directives and in the US by the AABB standards. The use of a statistical process control methodology is recommended in the related literature, including the EDQM guideline. The control reliability is dependent of the sampling. However, a correct sampling methodology seems not to be systematically applied. Commonly, the sampling is intended to comply uniquely with the 1% specification to the produced blood components. Nevertheless, on a purely statistical viewpoint, this model could be argued not to be related to a consistent sampling technique. This could be a severe limitation to detect abnormal patterns and to assure that the production has a non-significant probability of producing nonconforming components. This article discusses what is happening in blood establishments. Three statistical methodologies are proposed: simple random sampling, sampling based on the proportion of a finite population, and sampling based on the inspection level. The empirical results demonstrate that these models are practicable in blood establishments contributing to the robustness of sampling and related statistical process control decisions for the purpose they are suggested for. Copyright © 2017 Elsevier Ltd. All rights reserved.

Progress report on the k{sub infinity} measurements of HTGR type lattices with the oscillation technique at zero reactivity; Rapport d'advancement sur les mesures de K {sub infinite} des reseaux du type HTGR avec la technique de l'oscillateur a'reactivite' nulle

Energy Technology Data Exchange (ETDEWEB)

NONE

1971-01-15

The principles of measuring k {sub infinity} for a HTGR lattice using the oscillation technique with zero reactivity were already presented at the ''9th reactor physics meeting of countries participating in the Dragon project''. A brief summary of the essential characteristics of the experiment is followed by a status report on present work.

Mechanical design philosophy for the graphite components of the core structure of an HTGR

International Nuclear Information System (INIS)

Bodmann, E.

1987-01-01

Parallel to the layout and design of the graphite components for THTRs and the succeeding high temperature reactor projects, the design methods for graphite components have been improved over the years. The aim of this works is to develop the design methods which take into account both the particular properties of graphite and the particular functions of the components. Because of the close relation ship between materials and design codes, this development work has progressed with the development, testing and qualification of German reactor graphite. In this paper, the experience in this field of Hochtemperatur Reaktorbau GmbH and the results of the work and approach to the design problems are reported. The example of a HTR 500 design for a 550 MWe power station is taken up, and the core structure is explained. The graphite components are divided into three classes according to the stress limits. The loading of these components is reviewed. The aim of the design is not the complete avoidance of failure, but to avoid the failure of a single component from leading to a disadvantageous consequence which is not allowable. The classification of loading events, Weibull statistics and maximum allowable stress, the formation of the permissible stress, the assessment of stress due to multiaxial loading and so on are described. (Kako, I.)

[Sample preparation methods for chromatographic analysis of organic components in atmospheric particulate matter].

Science.gov (United States)

Hao, Liang; Wu, Dapeng; Guan, Yafeng

2014-09-01

The determination of organic composition in atmospheric particulate matter (PM) is of great importance in understanding how PM affects human health, environment, climate, and ecosystem. Organic components are also the scientific basis for emission source tracking, PM regulation and risk management. Therefore, the molecular characterization of the organic fraction of PM has become one of the priority research issues in the field of environmental analysis. Due to the extreme complexity of PM samples, chromatographic methods have been the chief selection. The common procedure for the analysis of organic components in PM includes several steps: sample collection on the fiber filters, sample preparation (transform the sample into a form suitable for chromatographic analysis), analysis by chromatographic methods. Among these steps, the sample preparation methods will largely determine the throughput and the data quality. Solvent extraction methods followed by sample pretreatment (e. g. pre-separation, derivatization, pre-concentration) have long been used for PM sample analysis, and thermal desorption methods have also mainly focused on the non-polar organic component analysis in PM. In this paper, the sample preparation methods prior to chromatographic analysis of organic components in PM are reviewed comprehensively, and the corresponding merits and limitations of each method are also briefly discussed.

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

Development of a system code for transient analysis in a HTGR

International Nuclear Information System (INIS)

Lee, Tae Beom

2004-02-01

A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

Conceptual study on HTGR-IS hydrogen supply system using organic hydrides

International Nuclear Information System (INIS)

Terada, Atsuhiko; Noguchi, Hiroki; Takegami, Hiroaki; Kamiji, Yu; Inagaki, Yoshiyuki

2012-02-01

We have proposed a hydrogen supply-chain system, which is a storage/supply system of large amount of hydrogen produced by HTGR-IS hydrogen production system. The organic chemical hydride method is one of the candidate techniques in the system for hydrogen storage and transportation. In this study, properties of organic hydrides and conventional hydrogen storage/supply system were surveyed to make use of the conceptual design of the hydrogen supply system using an organic hydrides method with VHTR-IS hydrogen production process (hydrogen production: 85,400 Nm 3 /h). Conceptual specifications of the main equipments were designed for the hydrogen supply system consisting of hydrogenation and dehydrogenation process. It was also clarified the problems of hydrogen supply system, such as energy efficiency and system optimization. (author)

LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR

Directory of Open Access Journals (Sweden)

S Simbolon

2015-03-01

Full Text Available LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR. The The aim of this research is to produce thousands of microsphere ADU (Ammonium Diuranate gels by using external gelation for an intermediate HTGR (High Temperature Gas-cooled Reactor nuclear fuel in laboratory-scale. Microsphere ADU gels were based on sol-solution which was made from a homogeneous mixture of ADUN (Acid Deficient Uranyl Nitrate which was containing uranyl ion in high concentration, a water soluble organic compound PVA (Polyvinyl Alcohol and THFA (Tetrahydrofurfuryl Alcohol. The simple unified home made laboratory experimental machine was developed to replace test tube experiment method which was once used due to a tiny amount of microsphere ADU gels produced. It consists of four main parts: tank filled sol-solution connecting to peristaltic pump and vibrating nozzle, preliminary gelation and gelation column. The machine has successfully converted 150 mL sol-solution into thousands of drops which produced 120 - 130 drops in each minute in steady state in ammonia gas free sector. Preliminary gelation reaction was carried out in ammonia gas sector where drops react with ammonia gas in a bat an eye followed by gelation reaction in column containing ammonia solution 7 M. In ageing process, ADU gels were collected and submerged into a vessel containing ammonia solution which was shaken for 1 hour in a shaker device. Isopropyl alcohol (90% solution was used to wash ADU gels and a digital camera was used to measured spherical form of ADU gels. Diameters in spherical spheroid form were found between 1.8 mm until 2.2 mm. The spherical purity of ADU gels were 10% - 20% others were oblate, prolate spheroid and microsphere which have hugetiny of dimples on the surface.   PRODUKSI GEL ADU SKALA LABORATORIUM DENGAN MENGGUNAKAN GELASI EKSTERNAL UNTUK BAHAN BAKAR ANTARA HTGR. Penelitian ini bertujuan untuk membuat ribuan gel bulat ADU (Ammonium

Promising materials for HTGR high temperature heat exchangers

International Nuclear Information System (INIS)

Kuznetsov, E.V.; Tokareva, T.B.; Ryabchenkov, A.V.; Novichkova, O.V.; Starostin, Yu.D.

1989-01-01

The service conditions for high-temperature heat-exchangers with helium coolant of HTGRs and requirements imposed on materials for their production are discussed. The choice of nickel-base alloys with solid-solution hardening for long-term service at high temperatures is grounded. Results of study on properties and structure of types Ni-25Cr-5W-5Mo and Ni-20Cr-20W alloy in the temperature range of 900 deg. - 1,000 deg. C are given. The ageing of Ni-25Cr-5W-5Mo alloy at 900 deg. - 950 deg. C results in decreased corrosion-mechanical properties and is caused by the change of structural metal stability. Alloy with 20% tungsten retains a high stability of both structure and properties after prolonged exposure in helium at above temperatures. The alloy has also increased resistance to delayed fracture and low-cycle fatigue at high temperatures. The developed alloy of type Ni-20Cr-20W with microalloying is recommended for production of tubes for HTGR high-temperature heat-exchangers with helium coolant. (author). 3 refs, 8 figs

Friction and wear behavior of Inconel 625 with Ni3Ti, TiN, TiC-CVD coatings in an HTGR environment

International Nuclear Information System (INIS)

Sarosiek, A.M.; Li, C.C.

1984-04-01

The following conclusions apply to Inconel 625 with Ni 3 Ti, TiN, TiC-CVD coatings, tested in an HTGR environment in a temperature range between 500 and 900 0 C at a contact pressure of 3.45 MPa. The average wear rate is very small varying between 0.0 and 1.7 x 10 -4 g/m. The wear rate shows little dependence on temperature and sliding velocity, increasing slightly as the temperature increases or as the sliding velocity decreases. Damage experienced by wear areas is minimal. Stick-slip friction was observed at low sliding velocity, however the friction coefficient is low (maximum 0.63) with an average value of about 0.44. The friction coefficient shows little dependence on temperature and sliding velocity, increasing slightly as the temperature increases, or as the sliding velocity decreases. Ni 3 Ti, TiN, TiC-CVD coatings, are considered effective in minimizing friction and wear damage of Inconel 625 in an HTGR environment

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.

OVERVIEW OF MODULAR HTGR SAFETY CHARACTERIZATION AND POSTULATED ACCIDENT BEHAVIOR LICENSING STRATEGY

Energy Technology Data Exchange (ETDEWEB)

Ball, Sydney J [ORNL

2014-06-01

This report provides an update on modular high-temperature gas-cooled reactor (HTGR) accident analyses and risk assessments. One objective of this report is to improve the characterization of the safety case to better meet current regulatory practice, which is commonly geared to address features of today s light water reactors (LWRs). The approach makes use of surrogates for accident prevention and mitigation to make comparisons with LWRs. The safety related design features of modular HTGRs are described, along with the means for rigorously characterizing accident selection and progression methodologies. Approaches commonly used in the United States and elsewhere are described, along with detailed descriptions and comments on design basis (and beyond) postulated accident sequences.

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

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

Present status of research on and development of HTGR techniques in the People's Republic of China

International Nuclear Information System (INIS)

Zhu Yongjun

1989-01-01

China is a developing country rich in coal, petroleum and hydropower resources. In the past ten years, energy production in China has had a large increase, but along with the development of economy, energy demands increase even more rapidly. Many problems exist in China's energy system. Considering the large energy demand in the near future and long-term energy strategy, China has already decided to develop nuclear power gradually. The first several nuclear power stations are being and will be built in the South-east sea shore region. Two 900 MW PWRs (from France) and one 300 MW PWR (home made) are now under construction at Daya Bay (Kwangton Province) and Qin Shan (Zhejiang Province). The succeeding PWR power plants are being planned. PWR nuclear power station has been selected for the beginning of China's nuclear power plan. For large scale utilization of nuclear power in the next century, the development of advanced reactor type with good safety and economy performances and high uranium utilization rate (uranium resources in China is not rich enough) is strategically important. HTGR, due to its inherent safety characteristics, high heat efficiency, flexible fuel system and wide application fields, is a prospective advanced reactor type. Research and development on HTGR have already been included in China's national technical development program and are going on smoothly

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

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

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

Probabilistic techniques using Monte Carlo sampling for multi- component system diagnostics

International Nuclear Information System (INIS)

Aumeier, S.E.; Lee, J.C.; Akcasu, A.Z.

1995-01-01

We outline the structure of a new approach at multi-component system fault diagnostics which utilizes detailed system simulation models, uncertain system observation data, statistical knowledge of system parameters, expert opinion, and component reliability data in an effort to identify incipient component performance degradations of arbitrary number and magnitude. The technique involves the use of multiple adaptive Kalman filters for fault estimation, the results of which are screened using standard hypothesis testing procedures to define a set of component events that could have transpired. Latin Hypercube sample each of these feasible component events in terms of uncertain component reliability data and filter estimates. The capabilities of the procedure are demonstrated through the analysis of a simulated small magnitude binary component fault in a boiling water reactor balance of plant. The results show that the procedure has the potential to be a very effective tool for incipient component fault diagnosis

HTGR molten salt sensible energy transmission and storage system design and costs

International Nuclear Information System (INIS)

1981-09-01

This report, which was prepared for Gas-Cooled Reactor Associates by United Engineers and Constructors under Contract No. GCRA/UE and C 81-203, presents the design and cost for a molten salt Sensible Energy Transmission and Storage (SETS) System. Although the reference system for this study is sized to be compatible with an 1170 MW(t) HTGR Nuclear Heat Source, the results and conclusions should be generally applicable to most large scale molten salt energy transmission system applications. A preliminary conceptual design is presented and alternative configurations are discussed. The sensitivity of system costs to variations in important system parameters are also presented. Costs for a reference case conceptual design are reported in constant 1980 dollars and inflated 1995 dollars

Waste heat gas utilization for HTGR gas turbine plant for sea water desalination

International Nuclear Information System (INIS)

Hunter, D.A.A.

1981-01-01

A thermodynamic analysis is performed for a HTGR - Gas Turbine Plant, coupled with a Rankine cycle for additional power generation and/or desalination of sea water with a multistage flash evaporator. Three basic alternatives are studied: a) Brayton cycle with inter-cooling and without regeneration, coupled with a Rankine cycle for power generation and steam for evaporator. b) Same as a) but without inter-cooling and with regeneration. c) Brayton cycle with regeneration, without inter-cooling, coupled with a Rankine cycle for sea water evaporator steam generation. The behavior of the three alternatives is established with a parametric study for the most representative variables. Economy, safety and control aspects were considered for the three different conceptions. (Author) [pt

HTGR power plant hot reheat steam pressure control system

International Nuclear Information System (INIS)

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

1975-01-01

A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

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

Computational model and performance optimization methodology of a compact design heat exchanger used as an IHX in HTGR; Modelo computacional y metodologia de optimizacion del funcionamiento de un intercambiador de calor de diseno compacto empleado como IHX en HTGR

Energy Technology Data Exchange (ETDEWEB)

De la Torre V, R.; Francois L, J. L., E-mail: delatorrevaldes@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico)

2017-09-15

The intermediate heat exchangers (IHX) present in high-temperature gas-cooled reactor (HTGR) present complex operating conditions, characterized by temperature values higher than 1073 K. Conventional designs of tubes and shell have shown disadvantages with respect to compact designs. In this work, computational models of a compact heat exchanger design, the printed circuit, were built under IHX conditions in a HTGR installation. In these models, a detailed geometry was considered in three dimensions, corresponding to a transfer unit of the heat exchanger. Computational fluid dynamics techniques and finite element methods were used to study the thermo-hydraulic and mechanical functioning of the equipment, respectively. The properties of the materials were defined as temperature functions. The thermo-hydraulic results obtained were established as operating conditions in the structural calculations. A methodology was developed based on the analysis of capital and operating costs, which takes into account the heat transfer, pressure drop and the mechanical behavior of the structure, in a single optimization variable. By analyzing the experimental results of other authors, a relationship was obtained between the operation time of the equipment and the maximum effort in the structure, which was used in the model. The results show that the model that allows a greater thermal efficiency differs from the one that has lower total cost per year. (Author)

1170-MW(t) HTGR-PS/C plant application study report: shale oil recovery application

International Nuclear Information System (INIS)

Rao, R.; McMain, A.T. Jr.

1981-05-01

The US has large shale oil energy resources, and many companies have undertaken considerable effort to develop economical means to extract this oil within environmental constraints. The recoverable shale oil reserves in the US amount to 160 x 10 9 m 3 (1000 x 10 9 bbl) and are second in quantity only to coal. This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to a shale oil recovery process. Since the highest potential shale oil reserves lie in th Piceance Basin of Western Colorado, the study centers on exploiting shale oil in this region

Effects of HTGR helium on the high cycle fatigue of structural materials

International Nuclear Information System (INIS)

Soo, P.; Sabatini, R.L.; Gerlach, L.

1982-01-01

High cycle fatigue tests have been conducted on Incoloy 800H and Hastelloy X in air and in HTGR helium environments containing low and high levels of moisture. For the helium environments, a higher mositure level usually gives a lower fatigue strength. For air, however, the strength is usually much lower than those for helium. For long test times at higher test temperatures, the fatigue strengths for Incoloy 800H often show a large decrease, and the fatigue limits are much lower than those anticipated from low cycle tests. Optical and scanning electron microscope observations were made to correlate fatigue life with surface and bulk microstructural changes in the material during test. Oxide scale cracking and spallation, surface recrystallization and intergranular attack appear to contribute to losses in fatigue strength

Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Kasten, P.R.; Bartine, D.E.

1981-01-01

A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Bartine, D.E.

1981-01-01

A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

Nanoscale imaging and identification of four-component carbon sample

Energy Technology Data Exchange (ETDEWEB)

Sheremet, Evgeniya S [ORNL; Rodriguez, Raul [Chemnitz University of Technology, Germany; Agapov, Alexander L [ORNL; Sokolov, Alexei P [ORNL; Hietschold, Michael [Chemnitz University of Technology, Germany; Zahn, Dietrich [Chemnitz University of Technology, Germany

2015-01-01

We demonstrate the unprecedented chemical imaging of individual constituents in a four-component sample made of several carbon allotropes: single-wall carbon nanotubes, graphene oxide, C₆₀ fullerene, and an organic residue. This represents a significant advance with respect to previous works that were mainly limited to systems with one or two components having very different chemical composition. Despite the spectral and spatial overlap from different components, plasmon-based nanospectroscopy allows the discrimination of all individual carbon nanomaterials here investigated. Among other physical insights such as doping observed in carbon nanotubes, the detailed chemical imaging of graphene oxide reveals higher defect concentration at the flake edges similarly to the case of graphene. We found that the organic residue has either low adsorption or lack of resonant enhancement on GO, in contrast to graphene, suggesting a decreased van der Waals interaction. Furthermore, this report paves the way for routine nanoscale analysis of complex carbon systems with spatial resolution of 15 nm and below.

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.

Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

Saurwein, John

2011-07-15

This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

International Nuclear Information System (INIS)

Saurwein, J.

2011-01-01

This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

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

Computational model and performance optimization methodology of a compact design heat exchanger used as an IHX in HTGR

International Nuclear Information System (INIS)

De la Torre V, R.; Francois L, J. L.

2017-09-01

The intermediate heat exchangers (IHX) present in high-temperature gas-cooled reactor (HTGR) present complex operating conditions, characterized by temperature values higher than 1073 K. Conventional designs of tubes and shell have shown disadvantages with respect to compact designs. In this work, computational models of a compact heat exchanger design, the printed circuit, were built under IHX conditions in a HTGR installation. In these models, a detailed geometry was considered in three dimensions, corresponding to a transfer unit of the heat exchanger. Computational fluid dynamics techniques and finite element methods were used to study the thermo-hydraulic and mechanical functioning of the equipment, respectively. The properties of the materials were defined as temperature functions. The thermo-hydraulic results obtained were established as operating conditions in the structural calculations. A methodology was developed based on the analysis of capital and operating costs, which takes into account the heat transfer, pressure drop and the mechanical behavior of the structure, in a single optimization variable. By analyzing the experimental results of other authors, a relationship was obtained between the operation time of the equipment and the maximum effort in the structure, which was used in the model. The results show that the model that allows a greater thermal efficiency differs from the one that has lower total cost per year. (Author)

Characterization of thermally sprayed coatings for high-temperature wear-protection applications

International Nuclear Information System (INIS)

Li, C.C.

1980-03-01

Under normal high-temperature gas-cooled reactor (HTGR) operating conditions, faying surfaces of metallic components under high contact pressure are prone to friction, wear, and self-welding damage. Component design calls for coatings for the protection of the mating surfaces. Anticipated operating temperatures up to 850 to 950 0 C (1562 to 1742 0 F) and a 40-y design life require coatings with excellent thermal stability and adequate wear and spallation resistance, and they must be compatible with the HTGR coolant helium environment. Plasma and detonation-gun (D-gun) deposited chromium carbide-base and stabilized zirconia coatings are under consideration for wear protection of reactor components such as the thermal barrier, heat exchangers, control rods, and turbomachinery. Programs are under way to address the structural integrity, helium compatibility, and tribological behavior of relevant sprayed coatings. In this paper, the need for protection of critical metallic components and the criteria for selection of coatings are discussed. The technical background to coating development and the experience with the steam cycle HTGR (HTGR-SC) are commented upon. Coating characterization techniques employed at General Atomic Company (GA) are presented, and the progress of the experimental programs is briefly reviewed. In characterizing the coatings for HTGR applications, it is concluded that a systems approach to establish correlation between coating process parameters and coating microstructural and tribological properties for design consideration is required

Oxidation parameters of nuclear graphite for HTGR air-ingress

International Nuclear Information System (INIS)

Kim, E.S.; No, H.C.

2004-01-01

In order to investigate chemical behaviors of the graphite during an air-ingress accident in HTGR, the kinetic tests on nuclear graphite IG-110 were performed in chemical reaction dominant regime. In the present experiment, inlet gas flow rate ranged between 8 and 18 SLPM, graphite temperatures and oxygen mole fraction ranged from 540 to 630degC and from 3 to 30% respectively. The test section was made of a quartz tube having 75 mm diameter and 750 mm length and the test specimen machined to the size of 21 mm diameter and 30 mm length was supported at the center of it by the alumina rod. The 15 kW induction heater was installed around the outside of test section to heat the specimen and its temperature was measured by 2 infrared thermometers. The oxidation rate was calculated from the gas concentration analysis between inlet and outlet using NDIR (non-dispersive infrared) gas analyzer. As a result the activation energy (Ea) and the order of reaction (n) were determined within 95% confidence level and the qualitative characteristics of the two parameters were also widely investigated by experimental and analytical methods. (author)

High-temperature irradiation effects on mechnical properties of HTGR graphites

International Nuclear Information System (INIS)

Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

1978-04-01

The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

Effect of microbiological testing on subsequent mid-infrared milk component analysis of the same milk sample.

Science.gov (United States)

Wojciechowski, Karen L; Melilli, Caterina; Barbano, David M

2014-09-01

Our objectives were to determine if mixing and sampling of a raw milk sample at 4°C for determination of total bacteria count (TBC) and if incubation at 14°C for 18h and sampling for a preliminary incubation (PI) count influenced the accuracy of subsequent fat, protein, or lactose measurement by mid-infrared (IR) analysis of milk from the same sample container due to either nonrepresentative sampling or the presence of microbial metabolites produced by microbial growth in the milk from the incubation. Milks of 4 fat levels (2.2, 3, 4, and 5%) reflected the range of fat levels encountered in producer milks. If the portion of milk removed from a cold sample was not representative, then the effect on a milk component test would likely be larger as fat content increases. Within the milks at each fat level, 3 treatments were used: (1) 20 vials of the same milk sampled for testing TBC using a BactoScan FC and then used for a milk component test; (2) 20 vials for testing TBC plus PI count followed by component test; and (3) 20 vials to run for IR component test without a prior micro sampling and testing. This was repeated in 3 different weeks using a different batch of milk each week. No large effect on the accuracy of component milk testing [IR fat B (carbon hydrogen stretch) and fat A (carbonyl stretch)] due to the cold milk sample handling and mixing procedures used for TBC was detected, confirming the fact that the physical removal of milk from the vial by the BactoScan FC (Foss Electric, Hillerød, Denmark) was a representative portion of the milk. However, the representativeness of any other sampling procedure (manual or automated) of a cold milk sample before running milk component testing on the same container of milk should be demonstrated and verified periodically as a matter of routine laboratory quality assurance. Running TBC with a BactoScan FC first and then IR milk analysis after had a minimal effect on milk component tests by IR when milk bacteria counts

HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

International Nuclear Information System (INIS)

Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.

1976-05-01

A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes

HTGR accident initiation and progression analysis status report. Volume 1. Introduction and summary

International Nuclear Information System (INIS)

Raabe, P.H.; Houghton, W.J.; Joksimovic, V.

1976-01-01

Probabilistic risk assessment techniques have been applied to obtain guidance in choosing nuclear safety research and development that is most worthwhile for high-temperature gas-cooled reactor (HTGR) nuclear power plants. The probabilistic techniques used are similar to those employed in the Reactor Safety Study for light water reactors (LWRs), WASH-1400, directed by Dr. N. C. Rasmussen. The recommendations for research include studies related to core heatup even though this event poses a very low risk to the public. In fact, it was found that under the many conditions covered by the study to date, even very infrequent accidents in HTGRs (say, once in ten million years) will not produce fatalities. Potential cost reduction areas have been found where alternate design options protect the public and meet regulatory safety criteria

Status of design code work for metallic high temperature components

International Nuclear Information System (INIS)

Bieniussa, K.; Seehafer, H.J.; Over, H.H.; Hughes, P.

1984-01-01

The mechanical components of high temperature gas-cooled reactors, HTGR, are exposed to temperatures up to about 1000 deg. C and this in a more or less corrosive gas environment. Under these conditions metallic structural materials show a time-dependent structural behavior. Furthermore changes in the structure of the material and loss of material in the surface can result. The structural material of the components will be stressed originating from load-controlled quantities, for example pressure or dead weight, and/or deformation-controlled quantities, for example thermal expansion or temperature distribution, and thus it can suffer rowing permanent strains and deformations and an exhaustion of the material (damage) both followed by failure. To avoid a failure of the components the design requires the consideration of the following structural failure modes: ductile rupture due to short-term loadings; creep rupture due to long-term loadings; reep-fatigue failure due to cyclic loadings excessive strains due to incremental deformation or creep ratcheting; loss of function due to excessive deformations; loss of stability due to short-term loadings; loss of stability due to long-term loadings; environmentally caused material failure (excessive corrosion); fast fracture due to instable crack growth

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

Clean power: for the nuclear future

International Nuclear Information System (INIS)

Nelken, H.H.; Barratt, R.O.; Mackenzie, S.J.; Essebaggers, J.; Gardner, W.A.; Walker, K.L.

1972-01-01

The activities of the Foster Wheeler Corporation in designing and supplying steam generators and other primary circuit components are reviewed. Descriptions are presented of design verification procedures, helium-heated steam generators for the HTGR, the LMFBR demonstration plant steam generator system, moisture separator/steam reheater units for water cooled reactors, and the layout of the Panama City, Florida facility for manufacture of HTGR components. (U.S.)

System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

McKellar, Michael G.; Harvego, Edwin A.; Gandrik, Anastasia A.

2010-01-01

A design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322 C and 750 C, respectively. The power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.

Initial Startup and Testing of the Fort St. Vrain HTGR - Lessons Learned which May Be Useful to the HTR-PM

International Nuclear Information System (INIS)

Brey, Larry H.

2014-01-01

Lessons Learned: Although the HTR-PM and FSV incorporate significant differences in their designs, there are lessons to be learned that are applicable to both plants. This is especially important for key systems that incorporate first-of-a-kind equipment. Basically, these lessons are just an application of common sense. • Complexity Breeds Unavailability. Incorporate system/components that are ruggedly simple in design with a history of reliable operation and minimal maintenance. • Assure Strong Expertise and Funding for this First HTR-PM. Quite likely, the successful startup and operation of this plant will require a level of support considerably greater than a typical nuclear plant. • Be Very Attentive to the Design Aspects of first-of-a-kind Components in the Class 1, Safety-Related Portions of the Plant. For example; a generic metallurgical failure could easily lead to a very long plant shutdown in order to redesign the failed component, re-license, manufacture, install and test prior to plant resuming plant operation. • Where Possible, Test all Key Systems/Components Prior to Installation using Actual Plant Configuration & Operating Characteristics This will help assure operational capability prior to application of nuclear heat. • Never Attempt to Start an Innovative Nuclear Power Plant Without First Having the Proper Regulatory Guides and Criteria in Place. FSV was licensed as a Research Facility. There was no Standard Review Plan or Regulatory Guides in place for the NRC (or PSC) to utilize in regulating this HTGR. • Do Not Be Reluctant to Incorporate a Generous Over-Build Capability into Systems/Components. It is significantly easier to design extra margin into the original compressors, pumps and motors than to be required to backfit into larger units after plant start-up. • Assure All Safety Documents Reflect the Actual Capability of the Plant to Respond to Accidents Described in the Safety Analysis. FSV was limited to 82% power during the

Apparatus and method for maintaining multi-component sample gas constituents in vapor phase during sample extraction and cooling

Science.gov (United States)

Felix, Larry Gordon; Farthing, William Earl; Irvin, James Hodges; Snyder, Todd Robert

2010-05-11

A dilution apparatus for diluting a gas sample. The apparatus includes a sample gas conduit having a sample gas inlet end and a diluted sample gas outlet end, and a sample gas flow restricting orifice disposed proximate the sample gas inlet end connected with the sample gas conduit and providing fluid communication between the exterior and the interior of the sample gas conduit. A diluted sample gas conduit is provided within the sample gas conduit having a mixing end with a mixing space inlet opening disposed proximate the sample gas inlet end, thereby forming an annular space between the sample gas conduit and the diluted sample gas conduit. The mixing end of the diluted sample gas conduit is disposed at a distance from the sample gas flow restricting orifice. A dilution gas source connected with the sample gas inlet end of the sample gas conduit is provided for introducing a dilution gas into the annular space, and a filter is provided for filtering the sample gas. The apparatus is particularly suited for diluting heated sample gases containing one or more condensable components.

Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

International Nuclear Information System (INIS)

Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

2014-12-01

The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V and V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement. (author)

Digital simulation of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant

International Nuclear Information System (INIS)

Ray, A.; Bowman, H.F.

1978-01-01

A nonlinear dynamic model of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant was derived in state-space form from fundamental principles. The plant model is 40th order, time-invariant, deterministic and continuous-time. Numerical results were obtained by digital simulation. Steady-state performance of the nonlinear model was verified with plant heat balance data at 100, 75 and 50 percent load levels. Local stability, controllability and observability were examined in this range using standard linear algorithms. Transfer function matrices for the linearized models were also obtained. Transient response characteristics of 6 system variables for independent step distrubances in 2 different input variables are presented as typical results

40 CFR 1065.145 - Gaseous and PM probes, transfer lines, and sampling system components.

Science.gov (United States)

2010-07-01

... Scrubber. You may use ammonia scrubbers for any or all gaseous sampling systems to prevent interference.... Follow the ammonia scrubber manufacturer's recommendations or use good engineering judgment in applying ammonia scrubbers. (f) Optional sample-conditioning components for PM sampling. You may use the following...

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)

Prevalence of Metabolic Syndrome Components in an Urban Mexican Sample: Comparison between Two Classifications

Science.gov (United States)

Isordia-Salas, Irma; Santiago-Germán, David; Rodrìguez-Navarro, Helem; Almaráz-Delgado, Martín; Leaños-Miranda, Alfredo; Anaya-Gómez, Francisco; Borrayo-Sánchez, Gabriela; Majluf-Cruz, Abraham

2012-01-01

Background. The aim of this study was to examine the prevalence of metabolic syndrome (MS) components in an urban Mexican sample. Methods. A total of 854 subjects were included. Anthropometric, blood pressure measurements, clinical data, and overnight fasting blood samples were obtained from all subjects. Results. In accordance with definitions by the American Heart Association/ National Heart, Lung, and Blood Institute (AHA/NHLBI) and the International Diabetes Federation (IDF), the prevalence of MS among participants was 59.7 and 68.7%, respectively. The prevalence of MS was higher in women and in individuals older than 45 years of age. More than 40% of the subjects fulfilled four criterions of MS according to both definitions. Conclusions. There was a high prevalence of MS components in an urban Mexican sample. Therefore, strong strategies had to be developed for early detection of MS and its components to prevent DMT2 and atherothrombotic complications in these patients. PMID:22203834

Operating experience of gas bearing helium circulators in HTGR development facility

International Nuclear Information System (INIS)

Shimomura, H.; Kawaji, S.; Fujisaki, K.; Ihizuka, T.

1988-01-01

The large scale helium gas test facility (HENDEL) has been constructed and operated since March 1982 at the Japan Atomic Energy Research Institute to develop HTGR components. The five electric driven gas circulators with dynamic gas bearings are used to circulate the helium gas of 4MPa and 400 deg. C in loops for their compactness, gas tightness, easy maintenance and free from gas contamination. All of these circulators are variable speed types of 3,000 to 12,000 rpm and have the same gas bearings and electric motors. The four machines among them are equipped with centrifugal impeller and one other machine has regenerative type, and the weight of both type rotors are nearly the same. After the troubles and repairing, both type of circulators were tested and the vibration characteristics were measured as preventing maintenance. From the test and measurements of the circulators, it was presumed that the first trouble on regenerative type was caused from excess unbalance force by falling off of a small pin from the rotating part and the second severe trouble on it was caused by the whipping in gas bearing. The static load on tilting pads indicated close relations to occurrence of the whirling through the measurements. It is recognized that fine balancing of the rotors and delicate clearance adjustment of the bearings are very important for the rotor stability and that the mechanism should be designed and machined so precise as to be adjustable. As the gas bearing would be damaged in an instantaneously short time, the monitoring technique for it should be so fast and predictive as to prevent serious damage. Through the tests, the vibration spectrum monitoring method seems to be predictive and useful for early detection of the shaft instability. It will be concluded that the gas bearing machine is an excellent system in its design philosophy, however, it also needs highly precise machining and delicate maintenance technique. 4 refs, 10 figs, 1 tab

Results of K_oo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

Energy Technology Data Exchange (ETDEWEB)

Benedetti, F; Brighenti, G.; Chiodi, P. L.; Ghilardotti, G.; Giuliani, C.

1974-10-15

This paper describes 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.

Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

International Nuclear Information System (INIS)

Tang, I.N.; Munkelwitz, H.R.

1982-07-01

The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated

Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

International Nuclear Information System (INIS)

1982-06-01

Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments

Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

Energy Technology Data Exchange (ETDEWEB)

1982-06-01

Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

Final summary report on the Peach Bottom End-of-Life Program

International Nuclear Information System (INIS)

Steward, K.P.

1978-07-01

HTGR design methods verifications have been performed under the Peach Bottom End-of-Life Program by comparison of actual with predicted physics, thermal, fission product, and materials behavior in Peach Bottom. These design methods verifications have utilized the data determined from nondestructive fuel and circuit gamma scanning on-site, from laboratory examinations of samples removed from the primary circuit, and from a complementary program of Peach Bottom fuel element postirradiation examinations at Oak Ridge National Laboratory. Fifty-five driver fuel elements were axially gamma scanned to determine fission product distributions for use in burnup calculations, power profile determinations, and fission product release and redistribution studies. The cesium plateout distribution in the primary circuit was subsequently mapped by gamma scanning the ducting at 12 locations, axially traversing 79 steam generator tubes with Cd Te detectors from the water side, and internally scanning two vertical runs of ducting. Component removal involved trepanning of the primary circuit ducting, obtaining access to the steam generator internals, and removing over 100 superheater, evaporator, and economizer tubing samples. During this phase, macroscopic examinations of the steam generator and ducting internals were performed. Subsequent laboratory examinations of removed samples included radiochemical tests to substantiate in-situ gamma scans and to determine 90 Sr distributions; metallurgical tests to evaluate surface films, microstructural changes, and residual mechanical properties; and tritium permeation tests to provide data for improving HTGR tritium release predictions. The condition and metallurgical integrity of all components examined were found to be excellent, with minimal coolant/substrate interaction and entirely acceptable residual mechanical properties. The suitability of the materials used in the construction of the Peach Bottom HTGR was thereby confirmed

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

Analysis of graphite dust deposition in hot gas duct of HTGR

International Nuclear Information System (INIS)

Peng Wei; Zhen Ya'nan; Yang Xiaoyong; Ye Ping

2013-01-01

The behavior of the graphite dust is important to the safety of high-temperature gas-cooled reactor (HTGR). The temperature field in hot gas duct was obtained using computational fluid dynamics (CFD) method. Further analysis to the thermo-phoretic deposition and turbulent deposition shows that as the dust particle diameter increases, the thermo-phoretic deposition efficiency decreases, and the turbulent deposition efficiency initially decreases and then increases. The comparisons of calculation results for two reactor powers, namely 30% FP (full power) and 100 % FP, indicate that the thermo-phoretic deposition efficiency is higher at 30% FP than that at 100% FP. while the turbulent deposition efficiency grows more rapidly at 100% FP. Besides, the results also demonstrate that the thermo-phoretic deposition and the turbulent deposition are nearly equivalent when particle sizes are small, while the turbulent deposition becomes dominant when particle sizes are fairly large. The calculation results by using the most probable distribution of particle size show that the total deposition of graphite dusts in hot gas duct is limited. (authors)

A contribution to the H.T.G.R. energy conversion issue

International Nuclear Information System (INIS)

Tilliette, Z.P.

1991-01-01

This paper reports on the HTGR temperature level which makes possible an energy conversion approach different from the common one using a steam cycle exclusively. By departing from the Rankine cycle practice or from the previously studied direct closed gas cycle, a combined gas-steam cycle is proposed, which follows the present trend in power generation. The cycle combination concept is particularly suitable to nuclear reactors because it maintains a relatively low reactor inlet temperature and features a helium (He) turbocompressor of a relatively low power level located on a secondary circuit, which leads to conventional operating conditions. The plant efficiency is significantly increased from 16 to 25 per cent, depending upon the reactor outlet temperature. Primary and secondary He mass flows are reduced and the reactor water ingress hazard is ruled out owing to a steam pressure slightly lower than the primary He one. An attractive arrangement can be proposed for both He primary heat exchanger and steam generator. The use of two turbosets can become an advantage in consequence of the actual possibility of operating only one of them

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 +131}I) 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.

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

ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

International Nuclear Information System (INIS)

Conklin, J.C.

2001-01-01

1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

Summary of ORNL high-temperature gas-cooled reactor program

International Nuclear Information System (INIS)

Kasten, P.R.

1981-01-01

Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

New mathematical method for the solution of gas-gas equilibria with special application to HTGR primary-coolant environments

International Nuclear Information System (INIS)

Bongartz, K.

1983-07-01

A new mathematical method and corresponding computer program have been developed that provide a general method for the numerical solution of an equilibrium problem involving the chemical interactions of gaseous species. The method and computer code were developed to calculate the equilibrium concentrations of impurity gases, such as CO, CO 2 , H 2 , H 2 O, CH 4 , and O 2 , which may be approached as the result of gaseous chemical reactions occurring within the hot primary coolant helium of a high-temperature gas-cooled reactor (HTGR). The method, however, can be applied to any gas mixture

Development of the krypton absorption in liquid carbon dioxide (KALC) process for HTGR off-gas reprocessing

International Nuclear Information System (INIS)

Glass, R.W.; Beaujean, H.W.R.; Cochran, H.D. Jr.; Haas, P.A.; Levins, D.M.; Woods, W.M.

1975-01-01

Reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel involves burning of the graphite-matrix elements to release the fuel for recovery purposes. The resulting off-gas is primarily CO 2 with residual amounts of N 2 , O 2 , and CO, together with fission products. Trace quantities of krypton-85 must be recovered in a concentrated form from the gas stream, but processes commonly employed for rare gas removal and concentration are not suitable for use with off-gas from graphite burning. The KALC (Krypton Absorption in Liquid CO 2 ) process employs liquid CO 2 as a volatile solvent for the krypton and is, therefore, uniquely suited to the task. Engineering development of the KALC process is currently under way at the Oak Ridge National Laboratory (ORNL) and the Oak Ridge Gaseous Diffusion Plant (ORGDP). The ORNL system is designed for close study of the individual separation operations involved in the KALC process, while the ORGDP system provides a complete pilot facility for demonstrating combined operations on a somewhat larger scale. Packed column performance and process control procedures have been of prime importance in the initial studies. Computer programs have been prepared to analyze and model operational performance of the KALC studies, and special sampling and in-line monitoring systems have been developed for use in the experimental facilities. (U.S.)

Horizon Expansion of Thermal-Hydraulic Activities into HTGR Safety Analysis Including Gas-Turbine Cycle and Hydrogen Plant

International Nuclear Information System (INIS)

No, Hee Cheon; Yoon, Ho Joon; Kim, Seung Jun; Lee, Byeng Jin; Kim, Ji Hwan; Kim, Hyeun Min; Lim, Hong Sik

2009-01-01

We present three nuclear/hydrogen-related R and D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the through flow calculation with a Newton- Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data

Sampling and chemical analysis of smoke gas components from the SP Industry Calorimeter

Energy Technology Data Exchange (ETDEWEB)

Maansson, M.; Blomqvist, P.; Isaksson, I.; Rosell, L.

1995-12-31

This report describes the sampling and chemical analyses of smoke gas components for combustion performed in the SP Industry Calorimeter, where continuous measurements of oxygen, carbon dioxide and carbon monoxide are an integrated part of the Calorimeter system. On-line measurements of nitrogen oxides and total amounts of unburnt hydrocarbons were performed. Hydrogen cyanide, hydrogen chloride and ammonia in the smoke were sampled and absorbed in impinger bottles and subsequently analyzed using wet chemical techniques. An adsorbent sampling system was designed to allow the identification and quantitative analysis of individual organic compounds in the smoke. Gas chromatography was utilized with a mass spectrometric detector for the identification and a FID for quantification of the total amounts as well as individual components. A procedure for cleaning the smoke gas duct in between the combustion experiments was designed and found to be effective. The materials studied were Nylon 66, polypropylene, polystyrene (with and without fire retardant), PVC, and chlorobenzene. A total of 19 large-scale tests were carried out. The mass of sample burnt ranged from 20 kg to 125 kg in an experiment. 14 refs, 11 tabs

Influence of the forces on the adhesion behavior of graphite dust in HTGR

International Nuclear Information System (INIS)

Peng Wei; Sun Xiaokai; Zhang Tianqi; Yu Suyuan

2015-01-01

The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. The present study focuses on the forces which make the graphite dust attach or detach from the surface in HTGR. The effect of graphite dust size, the fluid velocity and the surface energy between the particles and the substrate were investigated. The result showed that van der Waals adhesion force is the main factors affecting the dust attach on the surface, the gravity force and the electrostatic force were much smaller than it. For small particles, both the aerodynamic lift and drag are smaller than van der Waals adhesion force. While for the large particles, the coupled effects of aerodynamic lift and drag can make the dust detach from the substrate easier. Both the aerodynamic lift and drag forces will increase quickly as the fluid velocity increases. The surface energy is an important parameter for van der Waals adhesion force, which will decrease as the surface energy decreases. (author)

Numerical simulation of flow field in cooling tower of passive residual heat removal system of HTGR

International Nuclear Information System (INIS)

Li Xiaowei; Zhang Li; Wu Xinxin; He Shuyan

2011-01-01

Environmental wind will influence the working conditions of natural convection cooling tower. The velocity and temperature fields in the natural convection cooling tower of the HTGR residual heat removal system at different environmental wind velocities were numerically simulated. The results show that, if there is no wind baffle, the flow in the cooling tower is blocked when environmental wind velocity is higher than 6 m/s, residual heat can hardly be removed, and when wind velocity is higher than 9 m/s, the air even flow downwards in the tower, so wind baffle is very necessary. With the wind baffle installed, the cooling tower works well at the wind speed even higher than 9 m/s. The optimum baffle size and positions are also analyzed. (authors)

Creep properties of superalloys for the HTGR in impure helium environments

International Nuclear Information System (INIS)

Kawakami, H.; Nakanishi, T.

1981-01-01

This paper describes creep behaviors of two heat resistant alloys, Hastelloy X and Incoloy 800, in helium environments of the HTGR. In impure helium environments, these alloys are susceptible to carburization and oxidization. We have investigated these effects separately, and related them to the creep behaviors of the alloys. Experiments were carried out at 900 0 C both in helium and in air. Carburization results in decrease of secondary creep strain rate and delay of tertiary creep initiation. Oxidization caused decrease in tertiary creep strain rate of Hastelloy X, but did not that of Incoloy 800. Enhancement in tertiary creep strain rate of Hastelloy X in a very weakly oxidizing environment was confirmed in creep crack growth experiment using notched plate specimens. The rupture time of Hastelloy X in helium was short when compared with in air. Stress versus rupture time curves for both environments were parallel up to 5000 hours test, and a ratio of rupture stress in helium to that in air was about 0.9. In case of Incoloy 800, rupture time in helium was markedly prolonged as compared with that in air. (orig.)

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

Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

Directory of Open Access Journals (Sweden)

Kępisty Grzegorz

2015-09-01

Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

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

Development of high-strength concrete mix designs in support of the prestressed concrete reactor vessel design for a HTGR steam cycle/cogeneration plant

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.

1985-01-01

Design optimization studies indicate that a significant reduction in the size of the PCRV for a 2240 MW(t) HTGR plant can be effected through utilization of high-strength concrete in conjunction with large capacity prestressing systems. A three-phase test program to develop and evaluate high-strength concretes (>63.4 MPa) is described. Results obtained under Phase I of the investigation related to materials selection-evaluation and mix design development are presented. 3 refs., 4 figs

Experimental study on fundamental phenomena in HTGR small break air-ingress accident

International Nuclear Information System (INIS)

Kim, Jae Soon; Hwang, Jin-Seok; Kim, Eung Soo; Kim, Byung Jun; Oh, Chang Ho

2016-01-01

Highlights: • Air-ingress phenomena on the small break in a HTGR are experimentally investigated. • Experiment is investigated for various break sizes, angles, and density ratios. • Maximum air-ingress rate is observed at 120° in break angle. • This study reveals that air-ingress in the small break is governed by; buoyancy and flow inertia. • A non-dimensional parameter is newly proposed to determine the air-ingress flow regimes. • Newly proposed parameter is based on buoyancy versus inertia force. - Abstract: This study experimentally investigates fundamental phenomena in the HTGR small break air-ingress accident. Several important parameters including density ratio, break angle, break size, and main flow velocity are considered in the measurement and the analysis. The test-section is made of a circular pipe with small holes drilled around the surface and it is installed in the helium/air flow circulation loop. Oxygen concentrations and flow rates are recorded during the tests with fixed break angles, break sizes, and flow velocities for measurement of the air-ingress rates. According to the experimental results, the higher density difference leads to the higher rates of air-ingress with large sensitivity of the break angles. It is also found that the break angle significantly affects the air-ingress rates, which is gradually increased from 0° to 120° and suddenly decreased to 180°. The minimum air ingress rate is found at 0° and the maximum, at 110°. The air-ingress rate increases with the break size due to the increased flow-exchange area. However, it is not directly proportional to the break area due to the complexity of the phenomena. The increased flow velocity in the channel inside enhances the air-ingress process. However, among all the parameters, the main flow velocity exhibits the lowest effect on this process. In this study, the Froude Number relevant to the small break air-ingress conditions are newly defined considering both heavy

Assessment of effects of Fort St. Vrain HTGR primary coolant on Alloy 800. Final report

International Nuclear Information System (INIS)

Trester, P.W.; Johnson, W.R.; Simnad, M.T.; Burnette, R.D.; Roberts, D.I.

1982-08-01

A comprehensive review was conducted of primary helium coolant chemistry data, based on current and past operating histories of helium-cooled, high-temperature reactors (HTGRs), including the Fort St. Vrain (FSV) HTGR. A reference observed FSV reactor coolant environment was identified. Further, a slightly drier expected FSV coolant chemistry was predicted for reactor operation at 100% of full power. The expected environment was compared with helium test environments used in the US, United Kingdom, Germany, France, and Japan. Based on a comprehensive review and analysis of mechanical property data reported for Alloy 800 tested in controlled-impurity helium environments (and in air when appropriate for comparison), an assessment was made of the effect of FSV expected helium chemistry on material properties of alloy 800, with emphasis on design properties of the Alloy 800 material utilized in the FSV steam generators

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

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

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

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

A Principle Component Analysis of Galaxy Properties from a Large, Gas-Selected Sample

Directory of Open Access Journals (Sweden)

Yu-Yen Chang

2012-01-01

concluded that this is in conflict with the CDM model. Considering the importance of the issue, we reinvestigate the problem using the principal component analysis on a fivefold larger sample and additional near-infrared data. We use databases from the Arecibo Legacy Fast Arecibo L-band Feed Array Survey for the gas properties, the Sloan Digital Sky Survey for the optical properties, and the Two Micron All Sky Survey for the near-infrared properties. We confirm that the parameters are indeed correlated where a single physical parameter can explain 83% of the variations. When color (g-i is included, the first component still dominates but it develops a second principal component. In addition, the near-infrared color (i-J shows an obvious second principal component that might provide evidence of the complex old star formation. Based on our data, we suggest that it is premature to pronounce the failure of the CDM model and it motivates more theoretical work.

Applications of high-strength concrete to the development of the prestressed concrete reactor vessel (PCRV) design for an HTGR-SC/C plant

International Nuclear Information System (INIS)

Naus, D.J.

1984-01-01

The PCRV research and development program at ORNL consists of generic studies to provide technical support for ongoing PCRV-related studies, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities under this program have concentrated on the development of high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C plant, and the testing of models to both evaluate the behavior of high-strength concretes (plain and fibrous) and to develop model testing techniques. A test program to develop and evaluate high-strength (greater than or equal to 63.4 MPa) concretes utilizing materials from four sources which are in close proximity to potential sites for an HTGR plant is currently under way. The program consists of three phases. Phase I involves an evaluation of the cement, fly ash, admixtures and aggregate materials relative to their capability to produce concretes having the desired strength properties. Phase II is concerned with the evaluation of the effects of elevated temperatures (less than or equal to 316 0 C) on the strength properties of mixes selected for detailed evaluation. Phase III involves a determination of the creep characteristics and thermal properties of the selected mixes. An overview of each of these phases is presented as well as results obtained to date under Phase I which is approximately 75% completed

Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1984-01-01

Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

Chemical Components of Noncommercial Alcohol Beverage Samples: A Study With the Viewpoint of Toxic Components in Mashhad, Iran.

Science.gov (United States)

Dadpour, Bita; Hedjazi, Arya; Ghorbani, Hamideh; Khosrojerdi, Hamid; Vaziri, Seyed Mohsen; Malek Zadeh, Haleh; Habibi Tamijani, Amir

2016-06-01

Iran has one of the lowest alcoholic beverage use rates in comparison with other countries, because it is legally forbidden and because of religious beliefs. Even so, unrecorded and noncommercial alcohol remains a considerable concern, which needs special attention. In the current research, we have studied the general composition of noncommercial alcohol samples to identify potentially toxic components in the context of the city of Mashhad in IR Iran. Using a descriptive study, chemical composition records of alcohol samples obtained from Mashhad and its suburbs (from March 2013 to March 2014) were evaluated in terms of ethanol percentage and methanol percentage using gas chromatography. Likewise, the pH of the alcohol and the location of the sample were also considered. Some substances, such as inorganic elements, were not included because there was no information about these substances in the records. Of 877 reports of alcohol samples, more than 50% were obtained from Mashhad and the rest were from the suburbs. Of the reports, 57.5% were in the spring and summer, followed by 42.5% in the fall and winter. The mean (min-max) of ethanol percentage was 30.04% (0 - 98.4). In four cases, methanol was detected. The mean (min-max) of methanol percentage was 23% (4 - 95).The majority of the samples had an acidic pH. The composition of unrecorded samples did not raise major toxicological concern beyond ethanol in alcohol products. However, concentration levels of methanol in some unrecorded alcohol samples made these samples detrimental for human consumption.

Selection Component Analysis of Natural Polymorphisms using Population Samples Including Mother-Offspring Combinations, II

DEFF Research Database (Denmark)

Jarmer, Hanne Østergaard; Christiansen, Freddy Bugge

1981-01-01

Population samples including mother-offspring combinations provide information on the selection components: zygotic selection, sexual selection, gametic seletion and fecundity selection, on the mating pattern, and on the deviation from linkage equilibrium among the loci studied. The theory...

Development of analytical code `ACCORD` for incore and plant dynamics of High Temperature Gas-cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Itakura, Hirofumi

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The `ACCORD` code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the `ACCORD` code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Development of analytical code 'ACCORD' for incore and plant dynamics of High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko; Itakura, Hirofumi.

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The 'ACCORD' code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the 'ACCORD' code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

1972-10-15

The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

Safety considerations and countermeasures against fire and explosion at an HTGR-hydrogen production system. Proposal of safety design concept

International Nuclear Information System (INIS)

Nishihara, T.; Hada, K.; Shibata, T.; Shiozawa, S.

1996-01-01

Establishment of safety design concept and countermeasures against fire and explosion accidents is among key safety-related issues in an HTGR-hydrogen production system. We propose the different safety design concepts depending upon the origin of fire and explosion which may happen in the HTGR-hydrogen production plant. Against fire and explosion originated outside the reactor building (R/B), namely in the area of hydrogen production plant, the safety design concept is primarily to take a safe distance for preventing the damage on safety-related items or a proof wall if necessary. Because the hydrogen production plant is designed in the same safety level as a conventional chemical plant. The safe distance is proposed to limit an incident overpressure to 10 kPa so as not to suffer any damage on the items and to limit a wall-averaged temperature of concrete structures of the R/B to 175degC according to the current regulation. On the other hand, against a potential possibility of explosion originated inside the R/B, the safety design concept is to minimize the possibility of explosion low enough to assume no occurrence inside the R/B. That is, the measure is to exclude a simultaneous failure of a secondary helium piping and an endothermic chemical reactor. Furthermore, in severe accident condition in which the explosion may be postulated a priori, an incidental overpressure of explosion inside the reactor containment vessel (C/V) should be limited so as not to fail the C/V through restricting the amount of combustible gas ingress into the C/V by means of a combination of C/V isolation valve installed in the helium piping and emergency shut off valve in the process feed gas line. (author)

Medium-size high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

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

1980-08-01

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

Edge Principal Components and Squash Clustering: Using the Special Structure of Phylogenetic Placement Data for Sample Comparison

Science.gov (United States)

Matsen IV, Frederick A.; Evans, Steven N.

2013-01-01

Principal components analysis (PCA) and hierarchical clustering are two of the most heavily used techniques for analyzing the differences between nucleic acid sequence samples taken from a given environment. They have led to many insights regarding the structure of microbial communities. We have developed two new complementary methods that leverage how this microbial community data sits on a phylogenetic tree. Edge principal components analysis enables the detection of important differences between samples that contain closely related taxa. Each principal component axis is a collection of signed weights on the edges of the phylogenetic tree, and these weights are easily visualized by a suitable thickening and coloring of the edges. Squash clustering outputs a (rooted) clustering tree in which each internal node corresponds to an appropriate “average” of the original samples at the leaves below the node. Moreover, the length of an edge is a suitably defined distance between the averaged samples associated with the two incident nodes, rather than the less interpretable average of distances produced by UPGMA, the most widely used hierarchical clustering method in this context. We present these methods and illustrate their use with data from the human microbiome. PMID:23505415

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

A DATABASE FOR TRACKING TOXICOGENOMIC SAMPLES AND PROCEDURES WITH GENOMIC, PROTEOMIC AND METABONOMIC COMPONENTS

Science.gov (United States)

A Database for Tracking Toxicogenomic Samples and Procedures with Genomic, Proteomic and Metabonomic Components Wenjun Bao1, Jennifer Fostel2, Michael D. Waters2, B. Alex Merrick2, Drew Ekman3, Mitchell Kostich4, Judith Schmid1, David Dix1Office of Research and Developmen...

Matrix effects in PIXE evaluation of the major components in thick homogeneous samples

International Nuclear Information System (INIS)

Oystaeyen, B. van; Demortier, G.

1983-01-01

Matrix effects on the major component X-ray yields in PIXE measurements are studied in general terms with the Au-Cu-Ag matrix as practical example. Postulating firstly that all the major components may be simultaneously detected through one well-isolated peak for each of them, and secondly that a known reference material is available which contains all the elements of the unknown sample, we propose a direct method to extract the true concentrations of the latter taking into accout the matrix effects. The geometrical parameters describing the target location with respect to ion beam direction and detector position are also studied and a first attempt is made to include them in the data treatment. (orig.)

Two-dimensional vertical model seismic test and analysis for HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi; Honma, Toshiaki.

1983-02-01

The resistance against earthquakes of high-temperature gas cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. In the paper the test results of seismic behavior of a half-scale two-dimensional vertical slice core model and analysis are presented. The following results were obtained: (1) With soft spring support of the fixed side reflector structure, the relative column displacement is larger than that for hand support but the impact reaction force is smaller. (2) In the case of hard spring support the dowel force is smaller than for soft support. (3) The relative column displacement is larger in the core center than at the periphery. The impact acceleration (force) in the center is smaller than at the periphery. (4) The relative column displacement and impact reaction force are smaller with the gas pressure simulation spring than without. (5) With decreasing gap width between the top blocks of columns, the relative column displacement and impact reaction force decrease. (6) The column damping ratio was estimated as 4 -- 10% of critical. (7) The maximum impact reaction force for random waves such as seismic was below 60% that for a sinusoidal wave. (8) Vibration behavior and impact response are in good agreement between test and analysis. (author)

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)

Study on methodology to estimate isotope generation and depletion for core design of HTGR

International Nuclear Information System (INIS)

Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Shimakawa, Satoshi

2013-12-01

An investigation on methodology to estimate isotope generation and depletion had been performed in order to improve the accuracy for HTGR core design. The technical problem for isotope generation and depletion can be divided into major three parts, for solving the burn-up equations, generating effective cross section and employing nuclide data. Especially for the generating effective cross section, the core burn-up calculation has a technological problem in common with point burn-up calculation. Thus, the investigation had also been performed for the core burn-up calculation to develop new code system in the future. As a result, it was found that the cross section with the extended 108 energy groups structure from the SRAC 107 groups structure to 20 MeV and the cross section collapse using the flux obtained by the deterministic code SRAC is proper for the use. In addition, it becomes clear the needs for the nuclear data from an investigation on the preparation condition for nuclear data for a safety analysis and a fuel design. (author)

Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

Directory of Open Access Journals (Sweden)

Jianghai Li

2017-01-01

Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

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

Integrated and visual performance evaluation model for thermal systems and its application to an HTGR cogeneration system

International Nuclear Information System (INIS)

Qi, Zhang; Yoshikawa, Hidekazu; Ishii, Hirotake; Shimoda, Hiroshi

2010-01-01

An integrated and visual model EXCEM-MFM (EXergy, Cost, Energy and Mass - Multilevel Flow Model) has been proposed in this study to comprehensively analyze and evaluate the performances of thermal systems by coupling two models: EXCEM model and MFM. In the EXCEM-MFM model, MFM is used to provide analysis frameworks for exergy, cost, energy and mass four parameters, and EXCEM is used to calculate the flow values of these four parameters for MFM based on the provided framework. In this study, we used the tools and technologies of computer science and software engineering to materialize the model. Moreover, the feasibility and application potential of this proposed EXCEM-MFM model has been demonstrated by the example application of a comprehensive performance study of a typical High Temperature Gas Reactor (HTGR) cogeneration system by taking into account the thermodynamic and economic perspectives. (author)

Pressure vessel design codes: A review of their applicability to HTGR components at temperatures above 800 deg C

International Nuclear Information System (INIS)

Hughes, P.T.; Over, H.H.; Bieniussa, K.

1984-01-01

The governments of USA and Federal Republic of Germany have approved of cooperation between the two countries in an endeavour to establish structural design code for gas reactor components intended to operate at temperatures exceeding 800 deg C. The basis of existing codes and their applicability to gas reactor component design are reviewed in this paper. This review has raised a number of important questions as to the direct applicability of the present codes. The status of US and FRG cooperative efforts to obtain answers to these questions are presented

Gas-Cooled Thermal Reactor Program. Annual technical progress report for the period ending September 30, 1981

International Nuclear Information System (INIS)

1982-03-01

This report provides descriptions and results of the technical effort during FY81 on the Gas-Cooled Thermal Reactor Program. The FY81 work was organized according to the Work Breakdown Structure (WBS) for the National HTGR Program, and fell within five of the WBS tasks. The work on Market Definition and Development (WBS 03) was associated with estimating product costs for HTGR systems and their alternatives, projecting markets and market penetrations for these systems, and providing costs and market input to application analyses and component design. The Plant Technology (WBS 13) effort was mainly in the development of the systems dynamic computer code, STAR, for the transient analysis of HTGR's in reformer applications. The analysis of pebble bed reactors (PBR) was performed under Technology Transfer (WBS 15). The effort on components and systems within the nuclear heat source for reforming plants was performed under High Temperature Nuclear Heat Source (WBS 42)

Thermal transport properties of helium, helium--air mixtures, water, and tubing steel used in the CACHE program to compute HTGR auxiliary heat exchanger performance

International Nuclear Information System (INIS)

Tallackson, J.R.

1976-02-01

A description is presented of the thermal transport properties of the materials involved in digital computer calculations of heat transfer rates by the core auxiliary heat exchangers in large HTGR nuclear steam supply systems. These materials are pure helium, mixtures of helium with common gases having molecular weights in the range of 28 to 32, alloy steel tubing, and water. For use in programmed computations the viscosity, thermal conductivity, and specific heat are represented primarily by equations augmented by curves and tabulations. Materials supporting the development and selection of the property equations are included

Characterisation of blue-light stimulated luminescence components in different quartz samples: Implications for dose measurement

DEFF Research Database (Denmark)

Jain, M.; Murray, A.S.; Bøtter-Jensen, L.

2003-01-01

-aliquot dose-evaluation protocols seem to be relatively free of complications when applied to quartz dominated by the fast OSL component coming from 325degreesC TL region, but this may not be true for quartz in which other components are more significant. An adequate understanding of how different OSL...... results from measurement of (a) sensitisation, (b) thermal stability, (c) recuperation, and (d) infrared response as a function of stimulation temperature from 3 different samples of sedimentary quartz selected on the basis of relative OSL contribution from different blue light stimulated linearly...

Effect of heat source shape on the thermal field in the pebble bed core of High Temperature Gas-cooled Reactor (HTGR)

Energy Technology Data Exchange (ETDEWEB)

Chen, Leisheng; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

2015-10-15

In this study, in order to minimize the error brought by non-uniform heat flux, the spherical heaters are employed as heat source; subsequently, thermal field and heat transfer characteristics of the pebbles are investigated. The thermal field of the pebble surface in PBR is measured with heat source in different shapes. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTGR may affect the integrity of the pebbles, which has drawn the attention of many scientists to investigate the thermal field and to predict the maximum temperature locations in the pebbles using CFD method, Lee et.al has also done some experimental work on measuring the surface temperature of the pebbles as well as visualizing flow patterns of the coolant gas, and it was found that the temperature near the contacting points between pebbles was not higher than the flow stagnation points due to the higher thermal conductivity of the pebble. Certain error of temperature measurement might occur because of not very uniform heat flux in the pebbles since heater in cylindrical shape was utilized as heat source in previous experiment. More uniform heat flux and more complicated thermal profile are found in the result obtained using spherical heaters. The result shows that the temperature in contact point is higher than that in the top point, which is different from the previous results. The complex thermal phenomena observed in the lower-half side-sphere can be explained by the flow pattern near the surface.

Two-dimensional horizontal model seismic test and analysis for HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi; Honma, Toshiaki.

1988-05-01

The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)

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)

The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

International Nuclear Information System (INIS)

Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

1979-01-01

The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

Development of methane conversion improvement method by recycling of residual methane for steam reforming as a part of R and D of HTGR-hydrogen production system

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Haga, Katsuhiro; Aita, Hideki; Sekita, Kenji; Hino, Ryutaro; Koiso, Hiroshi.

1998-01-01

The purpose of the present study is to improve methane conversion for an HTGR-steam reforming system by recycling of residual methane. The residual methane in a product gas after steam reforming was recycled with a gas separator of polyimide membrane. Gas separation characteristics of the separator were investigated experimentally and numerically, and an experimental study on recycling system was carried out. The results showed that the recycling system improves apparent methane conversion, ratio of methane conversion to methane supply from a cylinder, from 20 to 32% compared with those without recycling. (author)

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

The materials programme for the high-temperature gas-cooled reactor in the Federal Republic of Germany: Status of the development of high-temperature materials, integrity concept, and design codes

International Nuclear Information System (INIS)

Nickel, H.; Bodmann, E.; Seehafer, H.J.

1990-01-01

During the last 15 years, the research and development of materials for high temperature gas-cooled reactor (HTGR) applications in the Federal Republic of Germany have been concentrated on the qualification of high-temperature structural alloys. Such materials are required for heat exchanger components of advanced HTGRs supplying nuclear process heat in the temperature range between 750 deg. and 950 deg. C. The suitability of the candidate alloys for service in the HTGR has been established, and continuing research is aimed at verification of the integrity of components over the envisaged service lifetimes. The special features of the HTGR which provide a high degree of safety are the use of ceramics for the core construction and the low power density of the core. The reactor integrity concept which has been developed is based on these two characteristics. Previously, technical guidelines and design codes for nuclear plants were tailored exclusively to light water reactor systems. An extensive research project was therefore initiated which led to the formulation of the basic principles on which a high temperature design code can be based. (author)

Bearing compartment seal systems for turbomachinery in direct-cycle HTGR power plants

International Nuclear Information System (INIS)

Adams, R.G.; Boenig, F.H.; Pfeifer, G.D.

1977-10-01

The direct-cycle High-Temperature Gas-Cooled Reactor (HTGR) employs a closed gas-turbine cycle with the primary reactor coolant (helium) as the working fluid. Design studies on this type of plant, carried out since 1971, have demonstrated, among other points, the advantages of the integrated arrangement, in which power from the cycle is transmitted to the electric generators by turbomachines completely enclosed in the reactor pressure vessel. A result of this arrangement is that the bearings are entirely enclosed within the primary coolant system of the reactor. An important aspect of the design of the turbomachinery is its prevention or minimization of the ingress of lubricants into the primary coolant system and its prevention of ingress of primary coolant into the bearing compartments. The design studies, which included thorough conceptual designs of the turbomachinery with emphasis on bearings and seals and their support systems showed that total exclusion of lubricant requires extremely complex seals and seal support systems. The variation of system low-end pressure with control actuation and the requirement that the bearing cavity pressure follow these variations were proved to further complicate the service system. The tolerance of even relatively minute amounts of entering lubricant during control transients will allow considerable simplification. This paper discusses the above-mentioned problems and their solutions in tracing the design evolution of a satisfactory bearing-compartment seals and service system. The resulting system appears to be feasible on the basis of experience with industrial gas turbines

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

Status of a reformer design for a modular HTGR in an in-line configuration

International Nuclear Information System (INIS)

Gluck, R.; Whitling, W.H.; Lipps, A.J.

1984-01-01

For the past several years the General Electric Company has had the technical lead on advanced concept studies for the Modular High Temperature Gas Cooled Reactor (HTGR) programs sponsored by the United States Department of Energy. The focus of the Modular Reactor System (MRS) effort is the development of a generic nuclear heat source capable of supplying heat to either a steam generator/electric cycle or a high temperature steam /methane reforming cycle. Some early ground rules for this study were that the reactor be designed for 950 deg. C direct cycle reforming and that the core be of the prismatic type. Since the prismatic core required control rods near the center of the core, the vertical in-line concept was selected to promote natural circulation cooling of the core for all potential transients except the depressurized core heatup transient. Although the requirement for a prismatic core has been eliminated for recent cost reduction studies, the vertical in-line configuration has been retained for its potential as the lowest cost configuration. This paper presents the results of recent design and analytical studies conducted to evaluate the feasibility of using a steam/methane reformer in a Vertical In-Line (VIL) arrangement with the generic nuclear heat source

Impact of sample size on principal component analysis ordination of an environmental data set: effects on eigenstructure

Directory of Open Access Journals (Sweden)

Shaukat S. Shahid

2016-06-01

Full Text Available In this study, we used bootstrap simulation of a real data set to investigate the impact of sample size (N = 20, 30, 40 and 50 on the eigenvalues and eigenvectors resulting from principal component analysis (PCA. For each sample size, 100 bootstrap samples were drawn from environmental data matrix pertaining to water quality variables (p = 22 of a small data set comprising of 55 samples (stations from where water samples were collected. Because in ecology and environmental sciences the data sets are invariably small owing to high cost of collection and analysis of samples, we restricted our study to relatively small sample sizes. We focused attention on comparison of first 6 eigenvectors and first 10 eigenvalues. Data sets were compared using agglomerative cluster analysis using Ward’s method that does not require any stringent distributional assumptions.

Stress analysis of two-dimensional C/C composite components for HTGR's core restraint techanism

International Nuclear Information System (INIS)

Satoshi Hanawa; Taiju Shibata; Jyunya Sumita; Masahiro Ishihara; Tatsuo Iyoku; Kazuhiro Sawa

2005-01-01

Carbon fiber reinforced carbon matrix composite (C/C composite) is one of the most promising materials for HTGRs core components due to their high strength as well as high temperature resistibility. One of the most attractive applications of C/C composite is the core restraint mechanism. The core restraint mechanism is located around the reflector block and it works to tighten reactor core blocks so as to restrict un-supposition flow pass of coolant gas (bypass flow) in the core. The restriction of bypass flow reads to the high efficiency of coolant flow rate inside of the reactor core. For the future HTGRs and VHTR (Very High Temperature Reactor), it is important to develop the core restraint mechanism with C/C composite substitute for metallic materials as used for HTTR. For the application of C/C composite to core restraint mechanism, it is important to investigate the applicability of C/C composite in viewpoint of structural integrity. In the present study, supposing the application of 2D-C/C composite to core restraint mechanism, thermal stress behavior was analyzed by considering the thickness of the C/C composite and the gap between reflector block and core restraint. It was shown from the thermal stress analysis that the circumferential stress decreases with increasing the gap and that the restraint force increases with increasing the thickness. By optimizing the thickness of C/C composite and gap between reflector block and core restraint, the C/C composite is applicable to the core restraint mechanism. (authors)

Evaluation of environmental samples containing heavy hydrocarbon components in environmental forensic investigations

Energy Technology Data Exchange (ETDEWEB)

Raia, J.C.; Blakley, C.R.; Fuex, A.N.; Villalanti, D.C.; Fahrenthold, P.D. [Triton Anal Corp, Houston, TX (United States)

2004-03-01

This article presents a procedure to evaluate and characterize environmental samples containing mixtures of hydrocarbons over a wide boiling range of materials that include fuels and other products used in commerce. The range of the method extends to the higher boiling and heavier molecular weight hydrocarbon products in the range of motor oil, bunker fuel, and heavier residue materials. The procedure uses the analytical laboratory technique of high-temperature simulated distillation along with mathematical regression of the analytical data to estimate the relative contribution of individual products in mixtures of hydrocarbons present in environmental samples. An analytical technique to determine hydrocarbon-type distributions by gas chromatography-mass spectrometry with nitric oxide ionization spectrometry evaluation is also presented. This type of analysis allows complex hydrocarbon mixtures to be classified by their chemical composition, or types of hydrocarbons that include paraffins, cycloparaffins, monoaromatics, and polycyclic aromatic hydrocarbons. Characteristic hydrocarbon patterns for example, in the relative distribution of polycyclic aromatic hydrocarbons are valuable for determining the potential origin of materials present in environmental samples. These methods provide quantitative data for hydrocarbon components in mixtures as a function of boiling range and 'hydrocarbon fingerprints' of the types of materials present. This information is valuable in assessing environmental impacts of hydrocarbons at contaminated sites and establishing the liabilities and cost allocations for responsible parties.

Status of the research and development at JAERI on the C/C composite control rod for HTGR

International Nuclear Information System (INIS)

Eto, M.; Ishiyama, S.; Ugachi, H.

1996-01-01

Control rod elements made of carbon-carbon composites were prepared and fracture-tested, aiming at the development of the more heat-resistant control rod which may impose the less restriction on the operation and shutdown of the HTGR. The control rod elements included pellet holder, lace truck and pin of PAN- or pitch-based composite material. On the basis of the results of fracture tests on the unirradiated elements, those made of PAN-based material were selected for an irradiation experiment. The irradiation was carried out in JRR-3 at 900 ± 50 deg. C to a maximum neutron fluence of 1 x 10 25 n/m 2 (E>29fJ). Fracture tests of the elements indicated that both fracture load and fracture displacement enough to assure the integrity of a control rod were maintained even after the irradiation. It was also found that both fracture strength and strain increased when applied load was parallel to the fiber felt plane, whereas the strength increase and strain decrease were observed for the load applied against the plane. (author). 11 refs, 16 figs

Possible applications and characteristics of HTGRs used for industrial co-generation

International Nuclear Information System (INIS)

Grebennik, V.N.

1984-01-01

The paper contains an overview of the HTGR's applications - high potential heat needs for industrial processes, covering a temperature range of approximately 500 to 1000 deg. C and higher are described, encompassing processes such as methane steam conversion, ammonia production, complex process of steam coal gasification, hydrogen production and coal hydrogenation for liquid synthetic fuel production, direct iron reduction etc. These needs are to be met by a reactor with power ranging from 1 to 10 GW(th). The USSR R and D work on high-temperature gas-cooled reactors is focused on the standardization of reactor components, high reliability and corresponding high NPP availability and safety. The main effort of the first experimental use of USSR HTGR units is concentrated on the steam conversion of methane. Other industrial processes will be studied at further stages of the HTGR development. (author)

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)

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973

International Nuclear Information System (INIS)

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

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

Quality analysis of commercial samples of Ziziphi spinosae semen (suanzaoren by means of chromatographic fingerprinting assisted by principal component analysis

Directory of Open Access Journals (Sweden)

Shuai Sun

2014-06-01

Full Text Available Due to the scarcity of resources of Ziziphi spinosae semen (ZSS, many inferior goods and even adulterants are generally found in medicine markets. To strengthen the quality control, HPLC fingerprint common pattern established in this paper showed three main bioactive compounds in one chromatogram simultaneously. Principal component analysis based on DAD signals could discriminate adulterants and inferiorities. Principal component analysis indicated that all samples could be mainly regrouped into two main clusters according to the first principal component (PC1, redefined as Vicenin II and the second principal component (PC2, redefined as zizyphusine. PC1 and PC2 could explain 91.42% of the variance. Content of zizyphusine fluctuated more greatly than that of spinosin, and this result was also confirmed by the HPTLC result. Samples with low content of jujubosides and two common adulterants could not be used equivalently with authenticated ones in clinic, while one reference standard extract could substitute the crude drug in pharmaceutical production. Giving special consideration to the well-known bioactive saponins but with low response by end absorption, a fast and cheap HPTLC method for quality control of ZSS was developed and the result obtained was commensurate well with that of HPLC analysis. Samples having similar fingerprints to HPTLC common pattern targeting at saponins could be regarded as authenticated ones. This work provided a faster and cheaper way for quality control of ZSS and laid foundation for establishing a more effective quality control method for ZSS. Keywords: Adulterant, Common pattern, Principal component analysis, Quality control, Ziziphi spinosae semen

HTGR accident initiation and progression analysis status report. Volume VII. Occupational radiation exposures from gas-borne and plateout activities

International Nuclear Information System (INIS)

1976-01-01

As a part of the Accident Initiation and Progression Analysis (AIPA) program, calculations were performed of the occupational dose rates and man-rem exposures from gas-borne and plateout activities in a reference 3000-MW(t) HTGR plant. The study included a preliminary survey to determine the most important contributors by operation or radiation source to the man-rem exposures. This survey was followed by detailed calculations for the most important cases. Median and 95 percent-confidence-level man-rem exposures per year were obtained for the gaseous activity in the containment building, moisture monitor system, analytic instrumentation, helium regeneration system, gas waste system, and reflector-block shipping. Median and 95 percent-confidence-level man-rem exposures per operation were obtained for the main-circulator removal, steam-generator tube plugging, and steam-generator removal and replacement. For each of these cases, the contributions to the man-rem exposures were calculated for the important isotopes

X-ray spectrum analysis of multi-component samples by a method of fundamental parameters using empirical ratios

International Nuclear Information System (INIS)

Karmanov, V.I.

1986-01-01

A type of the fundamental parameter method based on empirical relation of corrections for absorption and additional-excitation with absorbing characteristics of samples is suggested. The method is used for X-ray fluorescence analysis of multi-component samples of charges of welded electrodes. It is shown that application of the method is justified only for determination of titanium, calcium and silicon content in charges taking into account only corrections for absorption. Irn and manganese content can be calculated by the simple method of the external standard

Classification of soil samples according to their geographic origin using gamma-ray spectrometry and principal component analysis

International Nuclear Information System (INIS)

Dragovic, Snezana; Onjia, Antonije

2006-01-01

A principal component analysis (PCA) was used for classification of soil samples from different locations in Serbia and Montenegro. Based on activities of radionuclides ( 226 Ra, 238 U, 235 U, 4 K, 134 Cs, 137 Cs, 232 Th and 7 Be) detected by gamma-ray spectrometry, the classification of soils according to their geographical origin was performed. Application of PCA to our experimental data resulted in satisfactory classification rate (86.0% correctly classified samples). The obtained results indicate that gamma-ray spectrometry in conjunction with PCA is a viable tool for soil classification

Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

2006-12-15

The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

International Nuclear Information System (INIS)

Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

2006-12-01

The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

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)

High-temperature gas-cooled reactor: reformer application study

International Nuclear Information System (INIS)

1980-12-01

This Application Study evaluates the HTGR-R with a core outlet temperature of 850 0 C as a near-term Lead Project and as a vehicle to long-term HTGR Program objectives. The scope of this effort included evaluation of the HTGR-R technology, evaluation of potential HTGR-R markets, assessment of the economics of commercial HTGR-R plants, and the evaluation of the program scope and expenditures necessary to establish HTGR-R technology through the completion of the Lead Project. In order to properly assess the potential of the HTGR-R and the suitability of the HTGR-R as a Lead Project, additional work must be performed before a final judgment is rendered. Design trade-off studies and alternative applications must be investigated to determine if a commercial potential exists for the HTGR-R at 850 0 C. If commercial incentives are only identified for the HTGR-R with core outlet temperatures greater than 850 0 C, the design and development program duration and cost and the demonstration path for the HTGR-R must be reassessed

Effect of Binding Components in Complex Sample Matrices on Recovery in Direct Immersion Solid-Phase Microextraction: Friends or Foe?

Science.gov (United States)

Alam, Md Nazmul; Pawliszyn, Janusz

2018-02-20

The development of matrix compatible coatings for solid-phase microextraction (SPME) has enabled direct extraction of analytes from complex sample matrices. The direct immersion (DI) mode of SPME when utilized in conjunction with such extraction phases facilitates extraction of a wide range of analytes from complex matrices without the incurrence of fouling or coating saturation. In this work, mathematical models and computational simulations were employed to investigate the effect of binding components present in complex samples on the recovery of small molecules varying in logP for extractions carried out using the direct immersion approach. The presented findings corroborate that the studied approach indeed enables the extraction of both polar and nonpolar analytes from complex matrices, provided a suitable sorbent is employed. Further results indicated that, in certain cases, the kinetics of extraction of a given analyte in its free form might be dependent on the desorption kinetics of their bound form from matrix components, which might lower total recoveries of analytes with high affinity for the matrix. However, the binding of analytes to matrix components also enables SPME to extract a balanced quantity of different logP analytes, facilitated by multiphase equilibria, with a single extraction device.

Rapid Fractionation and Isolation of Whole Blood Components in Samples Obtained from a Community-based Setting.

Science.gov (United States)

Weckle, Amy; Aiello, Allison E; Uddin, Monica; Galea, Sandro; Coulborn, Rebecca M; Soliven, Richelo; Meier, Helen; Wildman, Derek E

2015-11-30

Collection and processing of whole blood samples in a non-clinical setting offers a unique opportunity to evaluate community-dwelling individuals both with and without preexisting conditions. Rapid processing of these samples is essential to avoid degradation of key cellular components. Included here are methods for simultaneous peripheral blood mononuclear cell (PBMC), DNA, RNA and serum isolation from a single blood draw performed in the homes of consenting participants across a metropolitan area, with processing initiated within 2 hr of collection. We have used these techniques to process over 1,600 blood specimens yielding consistent, high quality material, which has subsequently been used in successful DNA methylation, genotyping, gene expression and flow cytometry analyses. Some of the methods employed are standard; however, when combined in the described manner, they enable efficient processing of samples from participants of population- and/or community-based studies who would not normally be evaluated in a clinical setting. Therefore, this protocol has the potential to obtain samples (and subsequently data) that are more representative of the general population.

Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

2014-01-01

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

High temperature gas-cooled reactor: gas turbine application study

Energy Technology Data Exchange (ETDEWEB)

1980-12-01

The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

High temperature gas-cooled reactor: gas turbine application study

International Nuclear Information System (INIS)

1980-12-01

The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project