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Thursday, July 30, 2020 | History

2 edition of Core design study of a very high temperature reactor found in the catalog.

Core design study of a very high temperature reactor

K. E Asmussen

Core design study of a very high temperature reactor

by K. E Asmussen

  • 39 Want to read
  • 32 Currently reading

Published by The Office, for sale by the National Technical Information Service] in [Oakland, Calif.], [Springfield, Va .
Written in English

    Subjects:
  • High temperatures,
  • Nuclear pressure vessels -- Design and construction,
  • Gas cooled reactors

  • Edition Notes

    Statementby K. E. Asmussen and R. Rao ; ... prepared for the San Francisco Operations Office, Department of Energy
    SeriesGA-A ; 14586
    ContributionsRao, R., joint author, General Atomic Company, United States. Dept. of Energy. San Francisco Operations Office
    The Physical Object
    Pagination84 p. in various pagings :
    Number of Pages84
    ID Numbers
    Open LibraryOL14879490M

      Gas-cooled, graphite-moderated very-high-temperature reactors (VHTRs) are one design concept of Generation VI reactors. To examine the ability of the thermal hydraulics system code ATHLET for the safety assessment of these reactors, we performed simulations of the cores of both pebble bed and prismatic block VHTRs. The pebble bed high temperature gas-cooled reactor is a promising generation-IV reactor, which uses large fuel pebbles and helium gas as coolant. The pebble bed flow is a fundamental issue for both academic investigation and engineering application, e.g., reactor core design and safety analysis.

    A High Temperature Gas-cooled Reactor (HTGR) has several features different from conventional light water reactors such as inherent safety characteristics, high thermal efficiency and high economy. On the other hand, one of disadvantages of the HTGR with a prismatic core is to require rather long-term and expensive refueling, resulting in.   NUMERICAL ANALYSIS OF CORE THERMAL-HYDRAULIC FOR SODIUM-COOLED FAST REACTORS Alain CONTI1, Antoine GERSCHENFELD2, Yannick GORSSE2, 1: Reactor Studies Department, Nuclear Energy Division, CEA Cadarache, St Paul Lez Durance, France 2: System&Structure Modelling Department, Nuclear Energy Division, CEA .

    Neutronic studies of pin and pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core. The Very High Temperature Reactor (VHTR) is the leading candidate for the reactor component of the Next Generation Nuclear Plant (NGNP). This is because the VHTR demonstrates great potential in improving safety characteristics, being economically competitive, providing a high degree of proliferation resistance, and producing high outlet temperatures for .


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Core design study of a very high temperature reactor by K. E Asmussen Download PDF EPUB FB2

The very-high-temperature reactor (VHTR), or high-temperature gas-cooled reactor (HTGR), is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle.

The VHTR is a type of high-temperature reactor (HTR) that can conceptually have an outlet temperature of °C. The reactor core can be either a. Get this from a library. Core design study of a very high temperature reactor.

[K E Asmussen; R Rao; General Atomic Company.; United States. Department of Energy. San Francisco Operations Office.].

The energy released by fission events is transferred out of the reactor core via a working fluid. The defining characteristic of a VHTR is the very high temperature of this working fluid, capable of running an efficient power cycle or being used as a high temperature input for a chemical transformation process (e.g.

hydrogen production). The THTR (thorium high-temperature reactor ) of a pebble bed core design encountered technical problems after only a brief period of operation, and their scrutiny led to protracted shutdown.

The FSV and THTR were prematurely decommissioned largely as business by: 4. Abstract. A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to Cited by: The very-high-temperature reactor (VHTR) (see Fig.

) is a further step in the evolutionary development of high-temperature reactors (HTRs).The VHTR is a helium-gas-cooled, graphite-moderated, thermal-neutron-spectrum reactor with a core outlet temperature > °C, and a goal of °C, sufficient to support high-temperature processes such as production of.

Abstract. Neutronic studies of pin and pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met.

The Very High Temperature Reactor The term Very High Temperature Reactor (VHTR) loosely covers any reactor design with a coolant outlet temperature of °C or above.

The term typically refers to the next step in the evolutionary development of high-temperature gas-cooled reactors (HTGRs). The Next. Purchase Physics of High-Temperature Reactors - 1st Edition.

Print Book & E-Book. [email protected]{osti_, title = {Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October —December }, author = {Woods, Brian and Gutowska, Izabela and Chiger, Howard}, abstractNote = {Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor.

An innovative small transportable lead‐bismuth cooled fast reactor, named SPARK, with rated power of 20 MWth is proposed to operate for 20 years without refueling as a remote power supply. The SPARK core neutronics and thermal‐hydraulics design and preliminary safety analysis were performed in the current study.

CiteSeerX - Document Details (Isaac Councill, Lee Giles, Pradeep Teregowda): Neutronic studies of pin and pin stringer fuel assemblies have been performed to ascertain that core design requirements for the liquid-salt cooled Very High Temperature Reactor (LS-VHTR) can be met.

The stringer assemblies are being considered to eliminate the possibility of fuel. High Temperature Gas -cooled Reactor: Core Design Advanced Reactor Technologies Idaho National Laboratory.

Hans Gougar, PhD. Nuclear Engineer. Gerhard Strydom. National Technical Director – DOE Advanced Reactor Technologies Gas-Cooled Reactor Campaign. NRC HTGR Training July 16 The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [ MW(t)] fluoride-salt-cooled high-temperature reactor (FHR).

Core design of a direct-cycle, supercritical-pressure, light water reactor (SCLWR) is carried out. Double tube water rods are used for enhancing the moderation.

Whole coolant flows upward in the inner tube and downward in the outer tube of the water rod, and then the fuel is cooled. Conceptual design of the high temperature reactors cooled by supercritical water (SCR) is studied for 14 years at the University of Tokyo.

Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of. The high-temperature test reactor (HTTR) is a graphite-moderated gas-cooled research reactor in Oarai, Ibaraki, Japan operated by the Japan Atomic Energy uses long hexagonal fuel assemblies, unlike the competing pebble bed reactor designs.

HTTR first reached its full design power of 30 MW (thermal) in Other tests have shown that the core can reach. The key challenges in the reactor core design involve the selection of core geometry and materials capable of passively removing decay heat in loss of coolant events, achieving sufficiently high.

thermal reactor, equilibrium core, neutronic and thermal-hydraulic coupling, three-dimensional core calcula-tions, average coolant core outlet temperature I.

Introduction The concept of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) has been being developed at the University of Tokyo since It is a. inherently safe characteristics give high temperature reactors a significant advantage over other current reactor designs.

Other advantages of the pebble bed reactor can be found in the online capability of refueling, its possibility to achieve high fuel burnups and the high temperature of the heat produced the in core. was used in the core design to prevent heat transfer deterioration.

However, this condition limits the enthalpy rise in the core3). The removal of this condition from the core design criteria was a key technological breakthrough in designing high temperature core SCLWR-H and SCFR-H.

PCI (PCMI) is the limiting condition in LWR, because [email protected]{osti_, title = {Investigation on the Core Bypass Flow in a Very High Temperature Reactor}, author = {Hassan, Yassin}, abstractNote = {Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor .The present Very High Temperature Reactor (VHTR), operating at >C, Modeling of VHTR core and optimization of the VHTR core design was performed through the EVEN parity Transport code EVENT and the thermo-hydraulics code THERMIX.

The study was conducted in OECD Nuclear Science Committee (). 5. National initiatives and programmes.