Technology information on High Temperature Gas-cooled Reactor (HTGR)

HTGR Design Group, HTGR Project Management Office

JAEA-Technology 2023-019, 39 Pages, 2024/01

JAEA-Technology-2023-019.pdf:1.34MB

In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.

Conceptual study of ECH/ECCD system for fusion DEMO plant

Sakamoto, Keishi; Takahashi, Koji; Kasugai, Atsushi; Minami, Ryutaro; Kobayashi, Noriyuki*; Nishio, Satoshi; Sato, Masayasu; Tobita, Kenji

Fusion Engineering and Design, 81(8-14), p.1263 - 1270, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.10.010

no abstracts in English

Irradiation tests report of the 35th cycle in "JOYO"

*

JNC TN9440 2000-008, 79 Pages, 2000/08

JNC-TN9440-2000-008.pdf:2.33MB

This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).

An Evaluation study of measures for prevention of Re-criticality in sodium-cooled large FBR with MOX fuel

JNC TN9400 2000-038, 98 Pages, 2000/04

JNC-TN9400-2000-038.pdf:7.49MB

As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.

An Adjusted Cross Section Library for DFBR

Peter, J. Collins

PNC TN9410 97-034, 35 Pages, 1997/04

PNC-TN9410-97-034.pdf:1.07MB

While in the Reactor Physics Research Section of the Advanced Technology Division at OEC, I participated in the project to construct a data library for the demonstration fast breeder reactor (DFBR). This library would be produced using a combination of evaluated differential cross sections together with integral experimental data for fast reactors, so as to assure sufficiently accurate calculations for the DFBR designs. I had much experience of the design and use of experiments for the large-size cores at ZPPR under the title JUPITER which was performed under the USDOE/PNC joint agreement. My contribution here was mainly in extension of the experimental database to include the very-hard spectrum fast criticals from the Los Alamos National Laboratory (LANL). The data for these cores are described. Our work at ANLW with the GMADJ code, which is similar in effect to the ABLE code that we use at PNC, showed why many experiments are important in this project as well as those in the more obvious Pu/U oxide conventional cores which are of current interest for the DFBR. This point was not appreciated at PNC and is discussed here. The data from the fast spectrum critical experiments made at Los Alamos are described together with information that I have been able to find concerning the uncertainties. The main interest is these experiments has been for prediction of criticality. Consequently, the full covariance information that we would like has not been published. However, the uncertainty in the fuel content is, by far, the major contributor to the uncertainty. The LANL experiments have been a principal leg of the data testing for fast reactors for all versions of ENDF/B in the US. For our work, they provide measurements at Mev energies which are not available from the experiments in the softer-spectrum of the LMFBR.

None

Kato, Masato

PNC TN8600 94-005, 132 Pages, 1994/08

PNC-TN8600-94-005.pdf:7.95MB

no abstracts in English

None

*

PNC TN9360 93-001, 120 Pages, 1993/06

PNC-TN9360-93-001.pdf:5.08MB

no abstracts in English

None

*; *; Fukumura, Nobuo*; *; *; *; *

PNC TN1410 91-063, 239 Pages, 1991/08

PNC-TN1410-91-063.pdf:10.66MB

no abstracts in English

None

Seki, Masayuki

PNC TJ8009 91-001, 81 Pages, 1991/06

PNC-TJ8009-91-001.pdf:6.99MB

no abstracts in English