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JAEA Reports

A Note on the representation of rate-of-rise of the thermal stratification interface in reactor plenum

Tokuhiro, Akira; Kimura, Nobuyuki

JNC TN9400 2000-015, 26 Pages, 1999/09

JNC-TN9400-2000-015.pdf:1.43MB

The quantification of the rate-of-rise of the thermal stratification interface, a "thin" vertical zone where the temperature gradient is the steepest, is important in assessing the potential implications of thermally-induced stress problems in liquid-metal cooled reactors. Thermal stratification can likewise occur in confined volumes containing ordinary fluids (Pr1), where there is an input of thermal convective energy. In the prominent case of liquid metal reactors, there have been many studies on quantifying the rate-of-rise of a defined stratification interface, in terms of one or more of the following dimensionless groups, mainly: Richardson (Ri), Reynolds (Re), Grashof (Gr), Rayleigh (Ra) and/or Froude (Fr) numbers. Stratification is also a transient process in the volume in question. In the present work the anthors presents a derivation based on order-of-magnitude analysis (OMA), including an sensible energy balance, that produces a new representation more consistent than p

Journal Articles

Measurement of the power profile during nuclear excursions initiated by various reactivity additions using TRACY

; Nakajima, Ken; Yanagisawa, Hiroshi; ; ; *; *; Sakuraba, Koichi; Ono, Akio

Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 3, p.1277 - 1285, 1999/00

no abstracts in English

Journal Articles

Experiments and analyses on sodium void reactivity worth in mock-up cores of metallic fueled and MOX fueled fast reactors at FCA

Oigawa, Hiroyuki; Iijima, Susumu; Ando, Masaki

Journal of Nuclear Science and Technology, 35(4), p.264 - 277, 1998/04

https://doi.org/10.3327/jnst.35.264

Times Cited Count：6 Percentile：49.16(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

None

; ; Nogami, Yoshitaka; ; Seki, Masayuki; ;

PNC TN8410 96-198, 235 Pages, 1996/06

PNC-TN8410-96-198.pdf:11.35MB

None

JAEA Reports

Improvement of the level of safety for future FBR by means of passive safety features; 1:Assessment of passive safety measures and proposal of their R&D programs

; ; Uto, Nariaki; Yamaguchi, Akira; Kamide, Hideki; Ohshima, Hiroyuki; Hayashi, Kenji

PNC TN9410 94-235, 135 Pages, 1994/08

PNC-TN9410-94-235.pdf:6.67MB

In this report passive prevention and mitigation measures with regard to core disruptive accident in future large scale liquid metal cooled fast breeder reactors are discussed and assessed. First the criteria for the assessment of passive safety measures are proposed, and the commonly proposed passive prevention and mitigation measures are briefly reviewed. Then innovative prevention and mitigation measures are newly proposed to provide additional mechanisms to limit the core damage or to prevent a recriticality event during the core disruption process. After assessing these passive measures based on the proposed criteria, appropriate combinations of the measures are recommended. Further, required R&D programs to confirm their effectiveness are described including necessity of a new in-pile experimental program.

JAEA Reports

In-vessel thermohydraulic analysis of MONJU with AQUA (IV); Natural circulation analysis from a Full-power operation at the power ascension test period

Muramatsu, Toshiharu; *

PNC TN9410 92-106, 354 Pages, 1992/04

PNC-TN9410-92-106.pdf:26.19MB

A natural circulation analysis in the upper plenum of the MONJU reactor was conducted for transient simulating a pump coast down and reactor scram to a full-power operation condition using a multi-dimensional code AQUA. In the analysis, full options of the AQUA code (higher-order differencing schemes, an algebraic stress turbulence model, an adaptive Fuzzy control system, etc.) were used to obtain a refined numerical result. From the analysis, the following results have been obtained. (1)In a steady-state calculation simulating the full-power operation condition, maximum axial temperature gradient 154 $^{circ}$ C/m was calculated at the region between the upper and the lower flow holes. Therefore detailed measurements are necessary for thermal stress evaluation of internal components due to the axial temperature gradient at various power operation conditions. (2)In a transient caluculation simulating a natural circulation phenomenon, it was confirmed that a rising speed of the thermal stratification interface is delayed due to the decrease of a effective mixing volume in the upper plenum region. And the AQUA code calculated a discontinuity temperature transient (a hot shock continued from a cold shock) at the outlet nozzle of the reactor vessel due to the change of locally flow patterns in the upper plenum. Therefore it was concluded that detailed investigation is necessary using experimental data in various power operation conditions. (3)A gentle temperature transient was calculated with the AQUA code in comparison with a one-dimensional code. It is concluded that the one-dimensional code yields a conservative numerical result.

JAEA Reports

Analysis of the benchmark problem for the 7th meeting of the IAHR working group on advanced nuclear reactors thermal hydraulics with multi-dimensional code AQUA.

; *

PNC TN9410 91-217, 65 Pages, 1991/07

PNC-TN9410-91-217.pdf:1.09MB

For the purpose of the participation to a benchmark exercise of the 7th Meeting of the International Association for Hydraulic Research (IAHR) Working Group on Advanced Nuclear Reactors Thermal Hydraulics, which is to be held in Kernforschungszentrum Karlsruhe GmbH, FRG, August 27-29, 1991, two experimental cases of the benchmark problems were calculated using a combined method of higher-order accurate schemes and the Algebraic Stress turbulence Model (ASM) of a multi-dimensional themohydraulic analysis code AQUA developed at PNC. From the analyses, the following analytical results have been obtained: (1)Penetration phenomena at the inlet channel observed in experiments were predicted well with the combined method. (2)Axial distributions of temperature and horizontal velocity components in the test plenum agreed with experiments.

JAEA Reports

Investigation on efficiency of outer barrel for large-scale fast breeder reactor

; *; ; Yamaguchi, Akira; ; Sugawara, Satoru

PNC TN9410 91-089, 130 Pages, 1991/03

PNC-TN9410-91-089.pdf:5.14MB

In-vessel thermohydraulic analysis with multi-dimensional code AQUA was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to confirm efficiency of outer barrel equipments on a large-scale fast breeder reactor. Through the analysis using the AQUA code and the discussion based on their results, the following results have been obtained: [Main Loop Temperature Transient] The transient rate with the outer barrel equipments are approximately equal to the result when an inner barrel was adopted. [Thermal stratification] Axial temperature distributions are approximately equal to the result in the case without an inner barrel. Therefore appearance of an axial temperature distribution can be neglected from a structural design. [Circumferential Temperature Distribution] Maximum temperature gradient 104 $^{circ}$ C/m was confirmed. The value is equivalent to three times of that when an inner barrel was not adopted. Further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface Velocity] Maximum velocity is the same as that described for the case without an inner barrel. From the above results, it is concluded that the outer barrel considered here is an efficient equipment to relax the main loop temperature transient.

JAEA Reports

Investiation on presence of inner barrel for large fast breeder reactor

Muramatsu, Toshiharu

PNC TN9410 90-147, 115 Pages, 1990/10

PNC-TN9410-90-147.pdf:4.05MB

In-vessel thermohydraulics analysis was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to evaluate effects of an inner barrel on a large fast breeder reactor. Then four thermohydraulics phenomena, a thermal stratification, a main loop temperature transient, a circumferential temperature distribution and a sodium surface velocity were discussed. Through the analysis using the multi-dimensional code AQUA and the discussion, the following have been effects of the inner barrel as obtained: [Thermal Stratification] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of an axial temperature distribution can be neglected from a structural design. [Main Loop Temperature Transient] An inner barrel is required. Because a cold shock with maximum temperature transient -2.0 $^{circ}$ C/s occurred at a outlet nozzle when an inner barrel was not equipped. [Circumferential Temperature Distribution] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of the temperature distribution can be neglected from a structural design. But further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface velocity] An inner barrel is unnecessary. From the above results, it is concluded that an inner barrel is unnecessary if the cold shock is improved by a increase of effective mixing region on a design.

JAEA Reports

In-vessel thermohydraulic analysis of MONJU with AQUA code (III); Coolant mixing characteristics in the MONJU lower plenum

Muramatsu, Toshiharu

PNC TN9410 90-146, 64 Pages, 1990/10

PNC-TN9410-90-146.pdf:3.79MB

In vessel thermohydraulic analysis for coolant mixing characteristics of the MONJU lower plenum was carried out by multi-dimensional thermohydraulic analysis code AQUA. The characteristics comes up as an important problem when thermocouples located above the core utilized for a experimantally evalution of the MONJU in-vessel thermohydraulics. From the analysis, the following results have becn obtained: [Forced Convection Condition] (1)Coolant in the lower plenum was mixed efficiently bv a swiring flow effects. Therefore circumferential temperature distribution normalized by totally temperature difference at the high pressure plenum inlet region decreased to 20% from 50%. (2)In forced convetion condition, a buoyancy effect is smaller than inertia by the swirling flow effect. [Natural Convection Condition] Coolant mixing by the swirling flow cannot be expected due to increasing buoyancy effects. Therefore delicate cares on circumferential temperature distribution are necessary for in-vessel thermohydraulic evaluations.

JAEA Reports