Validation of feedback reactivity evaluation models for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04

https://doi.org/10.1115/1.4055328

Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.

Modelling concrete degradation by coupled non-linear processes

Oda, Chie; Kawama, Daisuke*; Shimizu, Hiroyuki*; Benbow, S. J.*; Hirano, Fumio; Takayama, Yusuke; Takase, Hiroyasu*; Mihara, Morihiro; Honda, Akira

Journal of Advanced Concrete Technology, 19(10), p.1075 - 1087, 2021/10

https://doi.org/10.3151/jact.19.1075

Concrete in a transuranic (TRU) waste repository is considered a suitable material to ensure safety, provide structural integrity and retard radionuclide migration after the waste containers fail. In the current study, coupling between chemical, mass-transport and mechanical, so-called non-linear processes that control concrete degradation and crack development were investigated by coupled numerical models. Application of such coupled numerical models allows identification of the dominant non-linear processes that will control long-term concrete degradation and crack development in a TRU waste repository.

Evaluation of feedback reactivity coefficients by inverse kinetics in Monju

Kitano, Akihiro; Nakajima, Ken*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04

The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

Critical analyses with ferromagnetic and plasma rotation effects and wall geometry for a high steady state tokamak

Kurita, Genichi; Bialek, J.*; Tsuda, Takashi; Azumi, Masafumi*; Ishida, Shinichi; Navratil, G. A.*; Sakurai, Shinji; Tamai, Hiroshi; Matsukawa, Makoto; Ozeki, Takahisa; et al.

Nuclear Fusion, 46(2), p.383 - 390, 2006/02

https://doi.org/10.1088/0029-5515/46/2/020

It is shown that critical beta is decreased by ferromagnetic effect by about 8% for /2, and denoting the permeability of ferromagnetic wall and vacuum, respectively, for tokamak of aspect ratio 3. The existence of the stability window for resistive wall mode opened by both effects of the toroidal plasma rotation and the plasma dissipation, which was not observed for high aspect ratio tokamak, is found for tokamak of aspect ratio 3. The effect of ferromagnetism on them is also investigated. The critical beta analyses of NCT (National Centralized Tokamak) plasma using VALEN code are started with stabilizing plate and vacuum vessel geometry with finite resistivity, and the results for passive effect of stabilizing plate are obtained. The calculations including stabilizing effect of the vacuum-vessel and also active feedback control are also performed for present design of NCT plasma.

External feedback effect on magnetic islands in tokamaks

Konovalov, S. V.; Mikhailovskii, A. B.*; Kovalishen, E. A.*; Kamenets, F. F.*; Ozeki, Takahisa; Shirokov, M. S.*; Takizuka, Tomonori; Tsypin, V. S.*

Doklady Physics, 49(7), p.405 - 408, 2004/07

https://doi.org/10.1134/1.1784852

Effect of feedback system, placed outside the vacuum vessel, on neoclassical tearing modes (NTMs) in tokamaks is analytically studied. External matching parameters for magnetic island chain are derived for the simplest case of step profile of equilibrium parallel current. The structure of dynamical equations for the island width and island rotation frequency in terms of matching parameters is explained. It is shown that such a feedback system can lead to essential shift of the island rotation, in particular, for stabilization of NTMs by the polarization current effect.

Experimental study on thermal-hydraulics and neutronics coupling effect on flow instability in a heated channel with THYNC facility

Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), 16 Pages, 2003/10

Thermal-hydraulic and neutronic dynamics are always interrelated in BWR core. This is called thermal-hydraulic and neutronic (T/N) coupling. Channel stability experiments with T/N coupling under non-nuclear condition are very limited. This is mainly due to the difficulties in the real-time simulation of neutron dynamics and in the fast-response void fraction measurement under high-pressure and temperature conditions. Authors have developed techniques to solve the above difficulties, and have succeeded in experimentally simulating T/N coupling under non-nuclear conditions with the THYNC facility. Using THYNC facility, T/N coupling effect on channel stability was investigated. Experiments were performed under Pressure=2-7MPa, Subcooling=10-40K, and Mass flux=270-660kg/ms. THYNC results indicated T/N coupling lowered the channel stability threshold. The reduction of channel stability threshold due to T/N coupling was small within 10% at 7MPa in the present THYNC experiment, although the experimental condition was set to be more severe than that supposed in a reactor.

Overview of JT-60U diagnostics for reactor-relevant plasma physics

Hatae, Takaki; JT-60 Team

Proceedings of 6th Japan-Australia Workshop on Plasma Diagnostics (CD-ROM), 16 Pages, 2002/00

The diagnostic system of JT-60U is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters such as electron temperature, electron density, ion temperature, plasma rotation, and plasma current (safety factor q) have been improved. As a result, the understanding of internal structure of plasmas has advanced. In particular, detailed studies of the internal transport barrier could be performed in reversed magnetic shear plasmas. Furthermore, a "current hole" or a nearly zero current density at the plasma center was discovered in JT-60U plasmas by using a motional stark effect (MSE) system. Several diagnostic signal have been linked to actuators such as neutral beam injectors, the gas feed system, and the electron cyclotron heating system establishing the real-time control of electron temperature, neutron yield, radiated power, and electron temperature gradient, and thus improving plasma performance.

Safety demonstration tests on thermal decomposition of nitrated solvent with nitric acid in nuclear fuel reprocessing plants (Contract research)

Tsukamoto, Michio; Takada, Junichi; Koike, Tadao; Watanabe, Koji*; Miyata, Teijiro*; Nishio, Gunji*; Murata, Mikio*; Uchiyama, Gunzo

JAERI-Tech 2001-031, 47 Pages, 2001/03

JAERI-Tech-2001-031.pdf:2.75MB

no abstracts in English

Study on reactivity feedback effects in the TRACY transient experiments

Nakajima, Ken; ;

Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 3, p.1286 - 1292, 1999/00

no abstracts in English