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Research and Development programs for HTGRs in JAERI

Abstract

Since 1969, JAERI has conducted research and development (R and D) programs for High-Temperature Gas-Cooled Reactors (HTGR). And the High Temperature engineering Test Reactor (HTTR), which will be the first High Temperature Gas Cooled Reactor (HTGR) in Japan, is under licensing process now. In this paper, some of the results of R and D are outlined in the following fields which are closely connected with the HTTR design, that is: i) fuel; ii) nuclear design; iii) thermal-hydraulic design; iv) graphite structure and v) high temperature metal structure. In the field of fuel, extensive investigations have been performed to develop the fabrication technology of coated particle fuel (cpf). In parallel, data of coated fuel particle failure and fission product release in in- and ex-reactor experiments as well as mechanical properties data were obtained and irradiation tests have been done using the Oarai Gas Loop No.1 (OGL-1) to verify the integrity of mass-produced fuel. Concerning the nuclear design, critical experiments were conducted using the Very High-Temperature Reactor Critical Assembly (VHTRC). Also carried out were hydrodynamical and thermal experiments using the Helium Engineering Demonstration Loop (HENDEL). On the graphite structures which compose the reactor internals, design criteria have been developed based on ASME  More>>
Authors:
Nishiguchi, Isoharu; Saito, Sinzo [1] 
  1. Department of HTTR Project, Japan Atomic Energy Research Institute (Japan)
Publication Date:
Jul 01, 1990
Product Type:
Conference
Report Number:
IAEA-TC-389.26
Reference Number:
EDB-01:068002
Resource Relation:
Conference: Technical committee meeting on gas-cooled reactor technology safety and siting, Dimitrovgrad (Russian Federation), 21-23 Jun 1989; Other Information: 18 refs, 15 figs, 5 tabs; PBD: 1990; Related Information: In: Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material, 724 pages.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; APPROPRIATE TECHNOLOGY; CREEP; DESIGN; FISSION PRODUCT RELEASE; GRAPHITE; HASTELLOY XR; HEAT EXCHANGERS; HELIUM; HTGR TYPE REACTORS; HTTR REACTOR; MATERIALS TESTING; MECHANICAL PROPERTIES; REACTOR COMPONENTS; REACTOR CORES; TEMPERATURE RANGE 1000-4000 K; THERMAL ANALYSIS; THERMAL FATIGUE
OSTI ID:
20180003
Research Organizations:
International Atomic Energy Agency, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
TRN: XA0101506036122
Availability:
Available from INIS in electronic form
Submitting Site:
INIS
Size:
page(s) 562-594
Announcement Date:

Citation Formats

Nishiguchi, Isoharu, and Saito, Sinzo. Research and Development programs for HTGRs in JAERI. IAEA: N. p., 1990. Web.
Nishiguchi, Isoharu, & Saito, Sinzo. Research and Development programs for HTGRs in JAERI. IAEA.
Nishiguchi, Isoharu, and Saito, Sinzo. 1990. "Research and Development programs for HTGRs in JAERI." IAEA.
@misc{etde_20180003,
title = {Research and Development programs for HTGRs in JAERI}
author = {Nishiguchi, Isoharu, and Saito, Sinzo}
abstractNote = {Since 1969, JAERI has conducted research and development (R and D) programs for High-Temperature Gas-Cooled Reactors (HTGR). And the High Temperature engineering Test Reactor (HTTR), which will be the first High Temperature Gas Cooled Reactor (HTGR) in Japan, is under licensing process now. In this paper, some of the results of R and D are outlined in the following fields which are closely connected with the HTTR design, that is: i) fuel; ii) nuclear design; iii) thermal-hydraulic design; iv) graphite structure and v) high temperature metal structure. In the field of fuel, extensive investigations have been performed to develop the fabrication technology of coated particle fuel (cpf). In parallel, data of coated fuel particle failure and fission product release in in- and ex-reactor experiments as well as mechanical properties data were obtained and irradiation tests have been done using the Oarai Gas Loop No.1 (OGL-1) to verify the integrity of mass-produced fuel. Concerning the nuclear design, critical experiments were conducted using the Very High-Temperature Reactor Critical Assembly (VHTRC). Also carried out were hydrodynamical and thermal experiments using the Helium Engineering Demonstration Loop (HENDEL). On the graphite structures which compose the reactor internals, design criteria have been developed based on ASME BandPV Code Section III Div.2, subsection CE and design data have been accumulated on a domestic graphite material. High temperature metal structure is also one of major subjects of R and D for HTGRs. Hastelloy XR, which is a modified version of Hastelloy X, was developed and various tests have been conducted which include creep tests, creep-fatigue tests, etc. to establish design criteria and allowables. Component tests of the Intermediate Heat Exchanger (IHX) have been also performed. (author)}
place = {IAEA}
year = {1990}
month = {Jul}
}