Basic study on the development of low exposure CT with Frisch grid avalanche diode, JAERI's nuclear research promotion program, H13-011 (Contract research)

Imanishi, Nobutsugu*; Ito, Akio*; Kanno, Ikuo*; Yoshida, Koji*; Onabe, Hideaki*

JAERI-Tech 2005-007, 45 Pages, 2005/03

JAERI-Tech-2005-007.pdf:6.84MB

no abstracts in English

The Development of MESHNOTE Code for Radionuclide Migration in the Near Field

; Makino, Hitoshi; Peter*

JNC TN8400 99-095, 69 Pages, 1999/12

JNC-TN8400-99-095.pdf:10.06MB

MESHNOTE code was developed to evaluate the engineered barrier system in collaboration with QuantiSci. This code is used to simulate glass dissolution, diffusive transport of nuclides in the buffer material and release to surrounding host rock. MESHNOTE is a one-dimensional finite difference, code, which uses cylindrical co-ordinates for the solution of a radially symmetric diffusion problem. MESHNOTE has the followig characteristics: (1) MESHNOTE can solve for diffusive transport of nuclides through an annulus shaped buffer region while accounting for multiple decay chains, linear and non-linear sorption onto the buffer materials and elemental solubility limits; (2) MESHNOTE can solve for ingrowth of plural daughter nuclides from a singular parent nuclide (branching), and the ingrowth of a singular daughter nuclide from plural parent nuclides (rejoining); (3) MESHNOTE can treat the leaching of nuclide from the vitrified waste and the release of nuclide from buffer to surrounding rock, which are boundary conditions for migration in the buffer, basing on the phenomena; (4) MESHNOTE can treat principal parameters (e.g. solubility and distribution coefficient) relevant to nuclide migration as time and space-dependence parameters; (5) The time stepping scheme in MESHNOTE is controlled by tolerance defined by the user. The time stepping will increase automatically while checking the accuracy of the numerical solution. The conceptual model, the mathematical model and the numerical implementation of the MESHNOTE code are described in this report and the characteristic functions of MESHNOTE are verified by comparing with analytical solutions or simulations produced with other calculation codes.

Development of whole core thermal hydraulic analysis code ACT; made based on several thermmal-hydraulic analysis codes; Code abstract and development of inter wrapper flow analysis program

Otaka, Masahiko; Ohshima, Hiroyuki

PNC TN9410 96-118, 26 Pages, 1996/04

PNC-TN9410-96-118.pdf:1.65MB

We have started to develop a whole core thermmal-hydraulic analysis code ACT(Analysis program of whole Core Thermal-hydraulics) for the purpose of evaluating detailed in-core thermal-hydraulic phenomena under various operation conditions, e,9., the normal operation and the transition from forced to natural circulation, of fast reactors. For the high accurate predictivity of the in-core thermal-hydraulics, key phenomena such as inter-wrapper flow (convection through the gaps between fuel subassemblies) and core-plenum thermal-hydraulic interaction should be accounted for. Therefore, ACT consists of four kinds of programs, i.e., intra-subassembly, inter-subassembly, upper plenum and primary loop (including intermediate heat exchanger) analysis programs, which will be made based on several thermal-hydraulic codes that have been developed at PNC and taken the verification and validation. The latter two programs are inevitable parts to give the proper boundary conditions of the in-core thermal-hydraulic analysis, especially in the natural circulation decay heat removal operation mode. These four programs will be coupled with each other and be calculated simultaneously by using parallel computers. In this report, the code development strategy and inter-wrapper flow analysis program which we developed as the first stage of the code development are presented. This program analyzes sodium single phase flow phenomena in inter-subassembly gap at whole core. The finite differential method is applied and the governing equations for fluid continuity, energy and momentum are solved simultaneously. The basic function of program was confirmed through the interwrapper flow analysis of a core consist of 37 fuel subassemblies. This program will be coupled with inter-subassembly analysis program at next stage.

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 154C/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.

The Effect of the virtual mass term on the stability of the two-fluid model against perturbations

Watanabe, Tadashi; Kukita, Yutaka

Nucl. Eng. Des., 135, p.327 - 340, 1992/00

https://doi.org/10.1016/0029-5493(92)90200-F

no abstracts in English

MONJU Shield plug gas blow-down test (II); Validation of COMMIX-PNC ver.MT and application to MONJU

*; Maekawa, I.*; Sato, Kazujiro*

PNC TN9410 87-056, 139 Pages, 1987/03

PNC-TN9410-87-056.pdf:15.95MB

The mass transport version (ver.MT) of three-dimensional thermal-hydraulic analysis code, COMMIX-PNC, has been developed to evaluate gas blow -down effects in the annulus between the MONJU closure head and plug port. The ver.MT has been validated through the analysis of the fundamental experiment of KCl transport and gas blow-down mock-up experiment. The fundamental experiments were carried out using a water cavity with 500 mm 500 mm 50 mm in size. The experiments began pouring KCl solution into the inlet of the cavity. The calculated histories of the KCl concentration transient agreed well with the experiment. For the mock-up gas blow-down experiment, three gas flowrate cases, 0.05m/min, 0.02m/min and 0.1m/min were calculated. Noble gas were predicted to reach the top part of the annulus only in the case with 0.1m/min flowrate. Through the application of the code to the MONJU configuration, the follwing have been effects of gas blow-down as obtained : [Normal gas blow condition] F.P. gases didn't enter into the annulus. [Half gas blow condition] The gases with 10% concentration reached the location of 3905 mm above the bottom of shielding plug, and [Gas blow trip condition] The gases with 13% concentration reached the door valve in 1000 sec. simulation. From the above results and their consistency with the evaluation of gas blow-down effects by the experimental correlation derived from the gas blow-down experiments, the correlation can be applicable to a complicated annulus like that of the MONJU.

Three Dimensional Thermal Analysis Code for Fuel Stock of VHTR -TBLOCK-

;

JAERI-M 85-145, 47 Pages, 1985/09

JAERI-M-85-145.pdf:1.03MB

no abstracts in English

Flow and Heat Transfer in Parallel Channel Attached with Equally-Spaced Ribs、3rd Report (Numerical Results in the Range of Reynolds Number from 150 to 850)

; Takizuka, Takakazu

JAERI-M 9564, 97 Pages, 1981/07

JAERI-M-9564.pdf:1.88MB

no abstracts in English

FLow and Heat Transfer in Parallel Channel Attached with Equally-Spaced Ribs; 1st.Report: Program for Numerical Calculation

; Takizuka, Takakazu

JAERI-M 8650, 49 Pages, 1980/01

JAERI-M-8650.pdf:1.22MB

no abstracts in English