Thermal-hydraulic design calculations for JRR-3 cold neutron source with the new moderator cell

Tokunaga, Sho; Horiguchi, Hironori; Nakamura, Takemi

JAEA-Technology 2023-001, 37 Pages, 2023/05

JAEA-Technology-2023-001.pdf:1.39MB

The cold neutron source (CNS) of the research reactor JRR-3 converts thermal neutrons generated in the reactor into low-energy cold neutrons by moderating them with liquid hydrogen stored in the moderator cell. Cold neutrons generated by the CNS are transported to experimental instruments using neutron conduits, and are used for many studies of physical properties, mainly in life science, polymer science, environmental science, etc. Improvement of cold neutron intensity is essential to maintain competitiveness with the world's research reactors in neutron science, and we are developing a new CNS that incorporates new knowledge. The current moderator cell for the CNS of JRR-3 is a stainless-steel container which is a canteen bottle type, and the cold neutron intensity can be improved by changing the material and shape. Therefore, the basic specifications of the new moderator cell were changed to aluminum alloy which has a smaller neutron absorption cross section, and the shape was optimized using a Monte Carlo code MCNP. Since these changes in specifications will result in changes in heat generation and heat transfer conditions, the CNS of JRR-3 was re-evaluated in terms of self-regulating characteristic, heat transport limits, heat resistance and pressure resistance, etc., to confirm its feasibility in thermal-hydraulic design. This report summarizes the results of the thermal-hydraulic design evaluation of the new moderator cell.

Reactor noise analysis for a graphite-moderated and -reflected core in KUCA

Sakon, Atsushi*; Nakajima, Kunihiro*; Takahashi, Kazuki*; Hohara, Shinya*; Sano, Tadafumi*; Fukaya, Yuji; Hashimoto, Kengo*

EPJ Web of Conferences, 247, p.09009_1 - 09009_8, 2021/02

https://doi.org/10.1051/epjconf/202124709009

In graphite-reflected thermal reactors, even a detector placed far from fuel region may detect a certain degree of the correlation amplitude. This is because mean free path of neutrons in graphite is longer than that in water or polyethylene. The objective of this study is experimentally to confirm a high flexibility of neutron detector placement in graphite reflector for reactor noise analysis. The present reactor noise analysis was carried out in a graphite-moderated and -reflected thermal core in Kyoto University Critical Assembly (KUCA). BF proportional neutron counters (1" dia.) were placed in graphite reflector region, where the counters were separated by about 35cm and 30cm -thick graphite from the core, respectively. At a critical state and subcritical states, time-sequence signal data from these counters were acquired and analyzed by a fast Fourier transform (FFT) analyzer, to obtain power spectral density in frequency domain. The auto-power spectral density obtained from the counters far from the core contained a significant degree of correlated component. A least-squares fit of a familiar formula to the auto-power spectral density data was made to determine the prompt-neutron decay constant. The decay constant was 63.314.5 [1/s] in critical state. The decay constant determined from the cross-power spectral density and coherence function data between the two counters also had a consistent value. It is confirmed that reactor noise analysis is possible using a detector placed at about 35cm far from the core, as we expected.

Research and development of low activation neutron absorber Au-In-Cd alloy at J-PARC Material and Life Science Experimental Facility

Oi, Motoki

Shiki, 43, P. 3, 2019/06

https://www.j-neutron.com/siki/Vol43.pdf

At the J-PARC Materials and Life Science Facility (MLF), we developed an Au-In-Cd alloy as a low activation neutron absorber for reducing the activity of moderator assembly, and put it to practical use in the reflector and moderator. As a method to confirm the distribution of indium in the alloy, pulse neutron imaging method is adopted, and the individual element distribution is measured by focusing on the resonance peak of indium, and the uniformity of the alloy is confirmed.

Current status of pulsed spallation neutron source of J-PARC

Takada, Hiroshi

JAEA-Conf 2017-001, p.51 - 56, 2018/01

https://doi.org/10.11484/jaea-conf-2017-001

A pulsed spallation neutron source of Japan Proton Accelerator Research Complex (J-PARC) is aimed at promoting a variety of cutting-edge materials researches at state-of-the-art neutron instruments with neutrons generated by a 3-GeV proton beam with a power of 1-MW at a repetition rate of 25 Hz. In 2015, for the first time it received 1-MW equivalent proton beam pulse, and the beam power for user program was ramped up to 500 kW. The moderator system of the neutron source was optimized to use (1) 100% para-hydrogen for increasing pulse peak intensity with decreasing pulse tail, (2) cylindrical shape with 14 cm diam. 12 cm long for providing high intensity neutrons to wide neutron extraction angles of 50.8 degrees, (3) neutron absorber made from Ag-In-Cd alloy to make pulse widths narrower and pulse tails lower. As a result, it gives highest intensity pulsed neutrons per incident proton in the world. Towards the goal to achieve the target operation at 1-MW for 5000 h in a year, efforts to mitigate cavitation damages at the target vessel front with injecting gas micro-bubbles into the mercury target are under way. Also, improvement of structural target vessel design is an urgent issue since there was failure twice at the water shroud of the mercury target due to the thermal stress during operating periods at 500 kW in 2015.

Effect of a particle diameter on the criticality of a MOX powder system

Takahashi, Satoshi*; Okuno, Hiroshi; Miyoshi, Yoshinori

JAERI-Tech 2005-056, 51 Pages, 2005/09

JAERI-Tech-2005-056.pdf:2.92MB

In the heterogeneous system of the mixed oxide fuel of uranium and plutonium, hereafter, MOX fuel, it was investigated whether the system could be modeled as a homogeneous system on the conditions which dealt with the MOX fuel of particle diameter 0.02mm or less in MOX fuel fabrication facilities in Japan. The infinite multiplication factor of the homogeneous system of the MOX fuel was first calculated, and the optimum moderation condition over the each ratio of PuO was determined. It was verified that carried out critical calculation for the heterogeneous system of the MOX fuel in which the spherical fuel diameter in a cube unit cell increased, and an atomic number ratio of hydrogen to heavy metal fixed conditions, and the probability for neutrons to escape resonance by a spherical fuel diameter no less than 0.1mm, and analyzed critical conditions etc. using a contiguous energy Monte Carlo code MVPII and JENDL3.3. The details of these calculations are reported. These results are expected to be quoted in a revised edition of "Nuclear Criticality Safety Handbook."

Summary report of the 7th Reduced-Moderation Water Reactor Workshop; March 5, 2004, JAERI, Tokai

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

JAERI-Conf 2005-009, 153 Pages, 2005/08

JAERI-Conf-2005-009.pdf:14.7MB

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI. The workshop on RMWRs is aiming at information exchange between JAERI and other organizations, and has been held every year since 1998. The program of the 7th workshop was composed of 5 lectures and an overall discussion time. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture as well as of the discussion time. In addition in Appendix, there are included presentation handouts of each lecture.

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 3; Analysis of liquid film falling down on inclined flat plate

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.25 - 31, 2005/03

https://doi.org/10.3327/taesj2002.4.25

no abstracts in English

3D measurement of void distribution of boiling flow in a tight-lattice rod bundle by neutron tomography

Kureta, Masatoshi; Tamai, Hidesada

Proceedings of 5th International Conference on Multiphase Flow (ICMF 2004) (CD-ROM), 10 Pages, 2004/06

3D void fraction distribution of boiling flow in a tight-lattice 7-rod bundle was measured by neutron radiography 3D computed tomography (neutron tomography) to investigate the flow characteristics in tight-lattice rod bundles and to verify the numerical analysis codes. The test section simulates the fuel rod bundle of the RMWR and consists of 7 heater rods with gap of 1.0mm and with diameter of 12.0mm. In this paper, the neutron tomography system, experiments and comparison of the measured data with a subchannel analysis code, COBRA-TF, are reported. It was found from this experiment that water layer which surrounds the heater rod becomes thick between rods, narrow region, and steam accumulates at the center region among three rods. COBRA-TF code overestimates the void fraction in a tight-lattice bundle compared with the present data.

Thermal-hydraulic analyses of poisoned cold moderator vessel, 1; Study on Poison Plate Layout

Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Teshigawara, Makoto; Hino, Ryutaro

JAERI-Tech 2004-018, 23 Pages, 2004/03

JAERI-Tech-2004-018.pdf:2.42MB

The Japan Atomic Energy Research Institute is constructing a mega-watt class spallation neutron source in cooperation with the High Energy Accelerator Research Organization. A cold moderator using liquid hydrogen is one of the key components in the system, which directly affects the neutronic performance both in intensity and pulse time structure. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant regions which would cause hot spots. A cold moderator with a poison plate (poisoned decoupled moderator) has a high possibility to generate the stagnant region on and near the poison plate. Thermal-hydraulic analyses were carried out with proposed inner structure of the poisoned cold moderator. The stagnant and recirculation regions could be reduced by making a gap between the poison plate end and the vessel bottom surface, and the local temperature rise also could be kept under the required design value.

Numerical analysis of two-phase flow characteristics in a reduced-moderation light water reactor

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada; Akimoto, Hajime

Transactions of the American Nuclear Society, 89, p.88 - 89, 2003/11

no abstracts in English