Development of experimental core configurations to clarify k variations by nonuniform core configurations

Gunji, Satoshi; Araki, Shohei; Suyama, Kenya

Nuclear Science and Engineering, 197(8), p.2017 - 2029, 2023/08

https://doi.org/10.1080/00295639.2022.2164151

The fuel debris generated by the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant is expected to have not only heterogeneous but also nonuniform compositions. Similarly, damaged fuel assemblies remaining in the reactor vessels also have nonuniform configurations due to some missing fuel rods. This non-uniformity may cause changing neutron multiplication factors. The effect of non-uniformity on the neutron multiplication factor is clarified by computations, and the possibility of experimentally validating the computations used for criticality management is being investigated. For this purpose, in this study the criticality effects of several core configurations of a new critical assembly, STACY, of the Japan Atomic Energy Agency with nonuniform arrangements of uranium oxide fuel rods, concrete rods, and stainless-steel rods were studied to confirm benchmarking potential. The difference in these arrangements changed the neutron multiplication factor by more than 1 $. We confirmed that changes in local neutron moderation conditions and the clustering of specific components caused this effect. In addition, the feasibility of benchmark experimental cores with nonuniform arrangements is evaluated. If benchmarking of such experiments could be realized, it would help to validate calculation codes and to develop criticality management methods by machine learning.

Generalized extreme value analysis of criticality tallies in Monte Carlo calculation

Ueki, Taro

Progress in Nuclear Energy, 159, p.104630_1 - 104630_9, 2023/05

https://doi.org/10.1016/j.pnucene.2023.104630

In this work, the methodology of Generalized Extreme Value (GEV) is applied to criticality tallies in Monte Carlo fission source cycles in order to evaluate the utility value of the distribution tail ends. Numerical results obtained under a sufficiently large number of particles per cycle show that the extreme value index (EVI) in GEV falls within the range of Weibull distribution including the EVI of Gumbel distribution as the role of a boundary value layer. GEV is also applied to a historically-challenging loosely-coupled system for demonstrating population diagnosis under an insufficient number of particles per cycle. It turns out that the transition from one equilibrium to other equilibrium makes the EVIs of upper and lower distribution tail ends depart from each other so that one of them falls in the range of Weibull distribution and the other in that of Frechet distribution.

Experimental study on velocity distribution in the subchannels of a fuel pin bundle with wrapping wire; Evaluation of the characteristics of flow field in 3-pin bundle

Hiyama, Tomoyuki; Aizawa, Kosuke; Nishimura, Masahiro; Kurihara, Akikazu

JAEA-Research 2021-009, 29 Pages, 2021/11

JAEA-Research-2021-009.pdf:2.25MB

In sodium-cooled fast reactors, high burnup of fuel is required for practical use. It is important to predict and evaluate the flow behavior in a fuel assembly because there is a concern that the heat removal capacity of the fuel assembly with high burnup will be locally reduced due to swirling and thermal deformation of the fuel rods. In this study, flow field measurement tests were conducted using a 3-pin bundle system test specimen for the purpose of elucidating the phenomenon and constructing a verification database for thermal hydraulics analysis code. The viewpoints of the experiment for elucidating the phenomenon are as follows; (1) Overall flow behavior in the subchannel including near the wrapping wire, (2) Relationship between Reynolds number including laminar flow region and flow field, and (3) Evaluation of the effect of the presence or absence of wrapping wire on the flow field. As a result, detailed flow field data in the subchannel was obtained by PIV measurement. It was found that when the wrapping wire crossed the subchannel, the flow occurred toward adjacent subchannel and the flow occurred that follows the winding direction of the wrapping wire. It was confirmed that the tendency of the flow velocity distribution of the Reynolds number in the laminar flow region is significantly different from that of the transition region and the turbulent region under the condition. The test was conducted using a same 3-pin bundle system without the wrapping wire, and it was confirmed that mixing by the wrapping wire occurred even in the laminar flow region.

A New critical assembly: STACY

Araki, Shohei; Gunji, Satoshi; Tonoike, Kotaro; Kobayashi, Fuyumi; Izawa, Kazuhiko; Ogawa, Kazuhiko

Proceedings of European Research Reactor Conference 2020 (RRFM 2020) (Internet), 7 Pages, 2020/10

Critical experiments of thermal neutron system are still expected to be playing an important role for wide technical issues. The Japan Atomic Energy Agency (JAEA) is renovating the Static Experimental Critical Facility (STACY) to maintain the experimental capability. The new STACY is designed as a general-purpose criticality facility. Its core mainly consists of low enriched UO fuel rods, grid plates, and light water moderator. The first experiment campaign in the new STACY aims to obtain criticality characteristics of fuel debris, which will be used in validation of criticality analysis methods. The designs of the experimental core configurations are in progress.

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

Atomistic simulation of phosphorus segregation to 3(111) symmetrical tilt grain boundary in -iron

Ebihara, Kenichi; Suzudo, Tomoaki

Modelling and Simulation in Materials Science and Engineering, 26(6), p.065005_1 - 065005_10, 2018/09

AA2018-0168.pdf:2.74MB

https://doi.org/10.1088/1361-651X/aace6a

Irradiation-induced grain boundary phosphorus segregation is an important factor for estimating the embrittlement of nuclear reactor pressure vessel steels, but the physical process of phosphorus migration to grain boundaries is still unclear. We numerically studied phosphorus migration toward 3(111) symmetrical tilt grain boundary in -iron using molecular dynamics. We found that, in the vicinity of the grain boundary within 1 nm distance, an iron-phosphorus mixed dumbbell and an octahedral interstitial phosphorus atom push a self-interstitial atom into the grain boundary, and the phosphorus atom becomes a substitutional atom. A phosphorus vacancy complex in the region also becomes dissociated, and the vacancy is absorbed in the grain boundary without dragging phosphorus. The results claim that a novel view of the segregation process is required.

Thermal-hydraulics analysis of fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08

A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.

Numerical analysis of flow field around simulated wire-wrapped fuel pins of fast reactor

Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08

In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.

Homogeneous transesterification kinetics of triglyceride to biodiesel using grafted fibrous catalyst

Ueki, Yuji; Seko, Noriaki

JAEA-Review 2013-059, JAEA Takasaki Annual Report 2012, P. 53, 2014/03

https://jopss.jaea.go.jp/pdfdata/JAEA-Review-2013-059.pdf

no abstracts in English

Annual report on operation, utilization and technical development of Hot Laboratories; From April 1, 2004 to March 31, 2005

Department of Hot Laboratories

JAERI-Review 2005-047, 95 Pages, 2005/09

JAERI-Review-2005-047.pdf:6.27MB

This is an annual report in 2004 fiscal year that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), and the Research Hot Laboratory (RHL) in the Department of Hot laboratories. In RFEF, BWR fuel rods were withdrawn from a fuel assembly irradiated for 5 cycles in the Fukushima-2 Nuclear Power Station Unit-1 and PIEs including nondestructive examination of those rods were carried out. In WASTEF, Slow Strain Rate Tests for detecting the susceptibility to IASCC, the corrosion test of reprocessing plant materials, tests for evaluating barrier performance in terms of waste disposal were performed. A secondary system pipe from the Mihama Nuclear Power Station Unit-3 was accepted to inspect the ageing fracture of it. In RHL, 15 lead cells are dismantled under the decommissioning plan at JAERI Tokai. And an arrangement of the RHL facility was started to use the storage of unirradiated nuclear materials.