Application of the modified neutron source multiplication method to the prototype FBR Monju

Truchet, G.*; Van Rooijen, W. F. G.*; Shimazu, Yoichiro*; Yamaguchi, Katsuhisa

Annals of Nuclear Energy, 51, p.94 - 106, 2013/01

https://doi.org/10.1016/j.anucene.2012.07.040

The Modified Neutron Source Method (MNSM) is applied to the prototype FBR Monju. Among Monju's particularities that have a big influence on the MNSM factors are: the presence of two californium sources near the core and the position of the detector, which is located far from the core outside of the reactor vessel. In order to evaluate the detector count rate, a propagation calculation has been conducted from the reactor vessel to the external detector. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). Results indicate a good agreement between the MNSM reactivity and the reactivity measured with other methods.

Gated multiple-sampling and tracking proportional chamber; New detector system for nuclear astrophysical study with radioactive nuclear beams

Hashimoto, Takashi; Ishiyama, Hironobu*; Ishikawa, Tomoko*; Kawamura, Takashi*; Nakai, Koji*; Watanabe, Yutaka*; Miyatake, Hiroari; Tanaka, Masahiko*; Fuchi, Yoshihide*; Yoshikawa, Nobuharu*; et al.

Nuclear Instruments and Methods in Physics Research A, 556(1), p.339 - 349, 2006/01

https://doi.org/10.1016/j.nima.2005.10.018

A new type of three dimensional tracking and proportional gas counter has been developed. Adopting a gating-grid system, performance of the detector becomes stable under the injection rate of charged particles less than 410 pps. It is a useful detection system for astrophysical experiments using radioactive nuclear beams, since the efficiency is so high as 100 %.

Key design parameter study (II) for large scale-up fast breeder reactor; Optimizing analysis of inherent negative reactivity feedback effect (I); Analysis on thermal transformation of core support plate

*; Tanigawa, Shingo*; *; Yamaguchi, Katsuhisa; *; *; *

PNC TN9410 88-141, 159 Pages, 1988/09

PNC-TN9410-88-141.pdf:10.2MB

The structural analyses of the core support plate have been applied to study thermal transfomation behaviors and the differences of the movement by changing analytical model, under anticipated transient without scram (ATWS) conditions of FBR. The analyses have been performed for 1000 MWe class loop type fast breeder reactor using a structural analysis code FINAS. The thermal-hydraulic results, which have been performed to ATWS conditions using a plant system code, were used as the thermal boundary conditions to the calculation. The scope of the analyses included a whole section of reactor vessel and the dead load of core assemblies was also considered. Following results were obtained from these studies. (1)The thermal transformation of a upper core support plate can be evaluated according to the free expansion behavior owing to the temperature change of core support plate itself. (2)The radial restriction due to core subassemblies has much influence on the axial bend of the core support plate. (3)There are some differences to the transformation results between by the whole model and by the one dimensional model during the thermal transient is large. Another analysis will be needed, however, about the reactivity change according to the displacement of the core structure.

Key Technological Design Study of a Large LMFBR(2); System Dyrawics Analysis for Mitigating ATWS Consequences of a 1000MWe Loop-Type LMFBR

Yamaguchi, Katsuhisa

PNC TN9410 87-161, 19 Pages, 1987/11

PNC-TN9410-87-161.pdf:2.49MB

A system dynamics analysis was applied to a 1000 MWe loop-type liquid-metal fast breeder reactor (IMFBR) to examine influence of possible innovative reactor designs on mitigating consequences of anticipated transients without scram (ATWS). Theanalysis included all the reactivity feedbacks having been employed in current analyses of hypothetical core disruptive accidents (HCDAs). In addition, the present analysis stressed inherent responses of the reactor system by including structural reactivity feedbacks due to axial expansion of control rod driveline (CRD) and radial expansion of reactor core driven by the expansion of the core support plate (CSP). An upper-core flow chimney was considered to make the CRD expansion feedback effective. The flow coastdown rate of the primary pump and the initial position of the control rod (CR) were treated as parameters. ATWS initiators examined were unprotected loss-of-flow (ULOF), loss-of-heat-sink (ULOHS) and transient overpower (UTOP). The ULOF accident was mitigated and peak sodium temperature was suppressed below boiling point by using the primary pump having a 40 s halving time of flow coastdown. The halving time could be shortened to 10 s by assuming that the CR was initially inserted into the active core by about 250 mm. The CRD expansion feedback controlled the earlier transient, and the CSP expansion feedback became dominant in the latter phase. The ULOHS consequence was eompletely enveloped in that of ULOF accident. The sodium temperatures in the primary system became lower than the ULOF case by about 100 C. The UTOP accident conceivable from the current plant design, i.e., the reactivity insertion of 60 ￠ with the rate of 1 to 3 ￠/s, suppressed sodium temperatures and fuel melt fractions below 650 C and 25 5, respectively.

Flow pattern and dryait under sodium boiling conditions at decary power levels

Yamaguchi, Katsuhisa; ;

Nuclear Engineering and Design, 99, p.247 - 263, 1987/02

https://doi.org/10.1016/0029-5493(87)90125-7

Quasi-steady low-heat-flux sodium boiling experiments were conducted examine the heat removal capability of liquid-metal fast breeder reactor(LMFBR) fuel subassemblies under sodium boiling conditions. The Sodium Boiling and Fuel Failure Propagation Test Facility, SIENA, and two 37-pin bundles, 37F and 37G, were used for this study.

Sodium Boiling Experiments at Decay Power Levels(4); Summary Assessment of the Low-Flow and Low-Heat Flax Sodium Boiling Expeiments at PNC

Yamaguchi, Katsuhisa; *; Aoki, Tadao

PNC TN941 85-56, 289 Pages, 1985/03

PNC-TN941-85-56.pdf:9.23MB

The objective of the Sodium Boiling Experiments at Decay Power Levels is to examine the heat removal capability of reactor fuel subassemblies under sodium boiling condition, which is a matter of arguments in analyzing the accidents like the loss of piping integrity and the loss of shutdown heat removal system. Prior to progressing the test program, a survey study was conducted to fix the scope within which an advanced investigation should be required to analyze the event sequence following after sodium boiling at low flow. The study focused on the results of the past low-flow and low-heat-flux boiling tests performed with the SIENA Facility, with special attension to summarizing the critical (dryout) conditions of the two-phase flow heat transfer. The corresponding data for water cases were also examined. The topical results are as follows: (1)The dryout phenomenon reproducible under annular flow condition at relatively high flow is well predicted to occur by the criterion that the exit quality equals to 0.5. (2)Even if the mismatching ratio of power to flow is increased at low flow range, the slug flow pattern is sustained, repeating the void expansion and contraction synchronized with the unstable flow oscillation. In this case, the extra-superior heat removal capability is expected due to strong heat sink around the voided region. The tendencies of the dryout quality data at annular flow on several parameters are resembled to those of water data, from which one can reach the conclusion that the dryout criterion confirmed here would be reasonable for the sodium flow cases. The forthcoming experiment should be, therefore, concentrated on examining the factors influential to the flow pattern transition and on generalizing the dryout data base at annular flow having less coolable nature.

Boiling and Dryout Conditions in Disturbed Cluster Geometry and Their Application to the Liquid-Metal Fast Breeder Reactor Local Fault Assessment

Yamaguchi, Katsuhisa; Nakamura, H.; Haga, Kazuo

Nuclear Science and Engineering, 88(3), p.464 - 474, 1984/11

https://doi.org/10.13182/NSE84-A18599

The effect of a local cooling disturbance caused by an edge-type blockage in a liquid-metal fast breeder reactor(LMFBR) fuel subassembly was investigated with a series of out-of-pile local blockage experiments with water and sodium. The heat exchange layer model first developed for central-type blockage cases applied well to the present edge-type cases. An empirical formula was developed for estimating maximum temperatures in various subassembly, and the conclusion was reached that a middle size edge-type blockage could lead to sodium boiling. The critical heat flux data of Power Reactor and Nuclear Fuel Development Corporation and Kernforschungszentrum Karlsruhe were correlated with the boiling inception heat flux for various core flow velociti

Experimental study of local sodium boiling detection by use of temperature and flow fluctuations at the outlet of subassembly

*; *; Haga, Kazuo*; Yamaguchi, Katsuhisa; *

PNC TN941 83-97, 71 Pages, 1983/06

PNC-TN941-83-97.pdf:2.53MB

Out-of-pile local boiling experiments were carried out with a wire spacer type 91-pin bundle. A half edge part of its cross-sectional area was blocked by a 5mm thick stainless-steel plate. The purpose of the experiments is to study the feasibility of detecting coolant boiling accident within a LMFBR fuel subassembly using the temperature and flow fluctuations at the outlet of individual subassembly. Initial conditions of the experiment are as follows: (1)Bundle inlet temperature : 400 500 C (2)Heat flux : 72 93 W/cm (3)Coolant flow velocity at the normal bundle section : 0.64 1.13 m/s. Local boiling was generated immediately behind the blockage by increasing heat flux gradually, with inlet sodium temperature and coolant flow rate being held to be constant. Specifications of the instrumentation system for measuring temperature and flow fluctuations are listed below: (1)Thermocouples Chromel-Alumel thermocouples, Grounded-type (O.D. 0.3mm) and ungrounded-type (O.D. 4.8mm) Time constant (63.2%) 10 msec and 2.14 sec respectively (2)Flowmeter Eddy-current type temperature and flowmeter Exciting frequency : 425 Hz (3)Fluctuation measuring circuit : Maximum gain : 60 dB Frequency characteristics : flat in the frequency range from 0.01 to 15 Hz, 20 dB/dec attenuation slope below 0.01 Hz and 200 dB/dec attenuation slope above 15 Hz. The following subjects have been studied. (1)To investigate how boiling informations are transferred. (2)Anomaly detection methods by use of the temperature and flow fluctuations. (3)Feasibility of detecting local boiling accident by fluctuation signals. These studies yielded the following conclusions. (1)Spectral peak arround 4 Hz in the power spectral density functions of the temperature fluctuations was observed at the downstream bundle section of the blockage. This frequency peak is consistent with that of void formation and collapse and may ...

Fuel failure and Failed fuel detection at "JOYO" and "MONJU"

Yamaguchi, Katsuhisa

no journal, , 

no abstracts in English

Study of thermal hydraulics and safety of FBR in Japan Atomic Energy Agency (JAEA)

Yamaguchi, Katsuhisa

no journal, , 

Comprehensive report of the safety study and its outcome will be made in the special committee on thermal-hydraulics and safety of FBR organized by Atomic Energy Society of Japan.