Evaluation of the minimum critical amount for heterogeneous lattice systems composed of fuel rods utilized in low-power water-moderated research and test reactors by using continuous-energy Monte Carlo code MVP with JENDL-4.0

Yanagisawa, Hiroshi

JAEA-Technology 2021-023, 190 Pages, 2021/11

JAEA-Technology-2021-023.pdf:5.25MB

Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.

Status of study of long-term assessment of transport of radioactive contaminants in the environment of Fukushima (FY2018) (Translated document)

Nagao, Fumiya; Niizato, Tadafumi; Sasaki, Yoshito; Ito, Satomi; Watanabe, Takayoshi; Dohi, Terumi; Nakanishi, Takahiro; Sakuma, Kazuyuki; Hagiwara, Hiroki; Funaki, Hironori; et al.

JAEA-Research 2020-007, 249 Pages, 2020/10

JAEA-Research-2020-007.pdf:15.83MB

The accident of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. occurred due to the Great East Japan Earthquake, Sanriku offshore earthquake, of 9.0 magnitude and the accompanying tsunami. As a result, large amount of radioactive materials was released into the environment. Under these circumstances, Japan Atomic Energy Agency (JAEA) has been conducting "Long-term Assessment of Transport of Radioactive Contaminants in the Environment of Fukushima" concerning radioactive materials released in environment, especially migration behavior of radioactive cesium since November 2012. This report is a summary of the research results that have been obtained in environmental dynamics research conducted by JAEA in Fukushima Prefecture.

Mock-up test of the modified STACY (Performance check of water feed and drain system)

Seki, Masakazu; Maekawa, Tomoyuki; Izawa, Kazuhiko; Sono, Hiroki

JAEA-Technology 2017-038, 52 Pages, 2018/03

JAEA-Technology-2017-038.pdf:4.6MB

The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY core is controlled by the water level fed in the core tank as well as the present STACY. In order to verify the basic design of the water feed and drain system of the modified STACY, we constructed a mockup test apparatus with almost the same structure and specifications as the modified STACY. In the mockup test, performance checks were pursued regarding limitation of maximum flow of water feeding, adjustment of the flow rate of water feeding, stop of water feeding and others. This report describes the outline and results of the mock-up test of the water feed and drain system of the modified STACY.

Data report of ROSA/LSTF experiment TR-LF-07; Loss-of-feedwater transient with primary feed-and-bleed operation

Takeda, Takeshi

JAEA-Data/Code 2016-004, 59 Pages, 2016/07

JAEA-Data-Code-2016-004.pdf:3.34MB

The TR-LF-07 test simulated a loss-of-feedwater transient in a PWR. A SI signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The primary feed-and-bleed operation was effective to core cooling because of no core uncovery.

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Present status of energy in Japan and HTTR project

Ogawa, Masuro; Nishihara, Tetsuo

Nuclear Engineering and Design, 233(1-3), p.5 - 10, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.07.018

An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20 % even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan. Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.

Report on investigation of cause of crack at instrumentation pipe in JMTR; Results of vibration and stress analysis

Hanawa, Satoshi; Tachibana, Yukio; Iyoku, Tatsuo; Ishihara, Masahiro; Ito, Haruhiko

JAERI-Tech 2003-064, 25 Pages, 2003/07

JAERI-Tech-2003-064.pdf:2.84MB

On the 147cycle operation, the water leakage was found at the pressure instrumentation pipe which is attached to the exit pipe of No.1 charge pump of the purification system of primary cooling system at JMTR in the Oarai establishment, JAERI. Then JMTR was shutted down manually on December 10th. It was predicted that the crack on the pressure instrumentation pipe was initiated and propagated by the cyclic load which was caused by the charge pump. Therefore, vibration and stress analyses of pressure instrumentation pipe were performed. From the vibration analysis, the natural frequency of the pressure instrumentation pipe of No.1 charge pump is between 5358Hz, which is close to the resonance frequency of 50Hz. From the stress analysis results, total stress generated on the pressure instrumentation pipe is 112.2MPa at the natural frequency of 53Hz and 74.2Mpa at 58Hz. It was found that the stress of 112.2MPa is close to the fatigue limit of used materials.

Chaotic behavior in a system simulating the pressure balanced injection system; Analysis of passive safety reactor behavior, JAERI's nuclear research promotion program, H12-012 (Contract research)

Madarame, Haruki*; Okamoto, Koji*; Tanaka, Gentaro*; Morimoto, Yuichiro*; Sato, Akira*; Kondo, Masaya

JAERI-Tech 2003-017, 156 Pages, 2003/03

JAERI-Tech-2003-017.pdf:5.31MB

no abstracts in English

Prevention measures for avoiding unexpected drifting of marine component in recovery equipment of significant metals from sea water; Positioning and monitoring system for marine component and improvement of its positioning accuracy

Tamada, Masao; Kasai, Noboru; Seko, Noriaki; Hasegawa, Shin; Kawabata, Yukiya*; Onuma, Kenji*; Takeda, Hayato*; Katakai, Akio; Sugo, Takanobu

JAERI-Tech 2001-065, 39 Pages, 2001/11

JAERI-Tech-2001-065.pdf:5.24MB

Positioning and monitoring system for marine component in recovery equipment of significant metals from seawater by using adsorbent was designed and assembled to avoid unexpected accidents. This system which was set on float part of the marine component obtains the position data by means of GPS satellites and send them to Takasaki and Mutsu establishments through satellite communication. As characteristic test for 20 days in the real sea, 256 data were obtained every 2 hours. The wrong data was eight and the positioning accuracy was 200.7m. To improve these characteristics, two new functions were added to the present firmware. These are to raise positioning resolutions in longitude and latitude from 0.001 to 0.00001 degree and to average remaining three data of positioning after maximum and minimum data were removed from continuous five times positioning. The improved system shows no wrong datum, average positioning of 6.74m. This performance is enough to prevent marine equipment from its drifting accident.

Vanadium uptake and an effect of vanadium treatment on F-labeled water movement in a cowpea plant by positron emitting tracer imaging system (PETIS)

Furukawa, Jun*; Yokota, Harumi*; Tanoi, Keitaro*; Ueoka, Shiori*; Matsuhashi, Shimpei; Ishioka, Noriko; Watanabe, Satoshi; Uchida, Hiroshi*; Tsuji, Atsunori*; Ito, Takehito*; et al.

Journal of Radioanalytical and Nuclear Chemistry, 249(2), p.495 - 498, 2001/08

https://doi.org/10.1023/A:1013276224300

We present real time Vanadate (V) uptake imaging in a cowpea plant by Positron Emitting Tracer Imaging System (PETIS). Vanadium-48 was produced by bombarding a Sc foil target with 50 MeV -particles at Takasaki Ion accelerators for Advanced Radiation Application (TIARA) AVF cyclotron. Then V was added to the culture solution to investigate the V distribution in a cowpea plant. The real time uptake of the V was monitored by PETIS. We measured the distribution of V in a whole plant after 3, 6 and 20 hours of V treatment by Bio-imaging Analyzer System (BAS). After the 20 hour treatment, vanadate was detected at the up-ground part of the plant. To know the effect of V uptake on plant activity, F-labeled water uptake was analyzed by PETIS. When a cowpea plant was treated with V for 20 hours before F-labeled water uptake experiment, the total amount of F-labeled water absorption was drastically desreased. Results suggest the inhibition of water uptake was mainly caused by the vanadate already moved to the up-ground part of the plant.