atw 2018-04v6

inforum

atw Vol. 63 (2018) | Issue 4 ı April

| | Fig. 7.

Pressure along two selected axial lines in the wire-wrapped rod bundle.

The inset specifies location of lines.

drop using data at corresponding

wire-wrap positions, i.e. from axial

positions 0.065 m to 1.268 m. The

mean pressure drop is 946 Pa/m.

2.3 Model validation

In this subsection, results of our

numerical study are compared to the

simplified Cheng and Todreas [1986]

correlation. The correlation was

recently recommended in (1) to

predict pressure drop (Δp) in bundles

with an accuracy of ±20 %. It applies

for a wide range of Reynolds- numbers.

The friction factor (f) is defined in

eq. 1, where d h,bdl , L, and u b

2

are

hydraulic diameter, length, and

average axial bundle velocity, respectively.

(1)

The correlation for f reads

for Re < Re L

for Re L ≤ Re ≤ Re T

for Re > Re T (2)

where

Re L = 300 x 10 1.7(P/D−1.0) (3)

Re T = 10,000 x 10 0.7(P/D−1.0) (4)

ψ = log(Re/Re L ) / log(Re T /Re L ) (5)

C fL = (-974.6 + 1612.0(P/D) −

598.5(P/D) 2 )(H/D) .06-0.085(P/D)

(6)

C fT = (0.8063 − 0.9022(log(H/D)) +

0.3526(log(H/D)) 2 ) ×

(P/D) 9.7 (H/D) 1.78-2.0(P/D) (7)

We compare the nominal flow case

of 3.580 kg/s which corresponds to a

velocity of 0.2 m/s and Re is 8910,

which is in the transient region.

According to eqns (3) and (4), Re L and

Re T are 902 and 15735, respectively.

The calculated friction factor f

equates to 0.0557. This corresponds

to a pressure drop in the bundle of

1407.2 Pa. The predicted pressure

drop resulting from the CFD study is

1,138 Pa. The difference is near 19 %,

which lays within the accuracy limits.

In future thermos-hydraulic simulations,

the current model can be

applied. For posttest analysis, additional

sensitive studies might be

necessary to further reduce the

uncertainty.

Conclusions

The flow in the gap between neighbouring

fuel assemblies plays an

important role in transients between

forced and natural convection. At

KALLA an experiment on the interwrapper

flow is currently setup and

accompanied by pre-test numerical

CFD studies. These proof that both

the flow in the gap region and the

fuel bundle are not influenced by the

upstream flow-conditioning region.

Moreover, development length are

much shorter than the unheated

length of the test section, so that

the thermal field is uninfluenced by

flow non-uniformities. Preliminary

comparison of pressure losses computed

by CFD and correlation provide

reasonable agreement for both the

gap and bundle. The result of our

study enters pre-test studies of the

thermal field within the EU-H2020

SESAME project. There complete

simulation of the test section consisting

of three bundles connected

by the gap region including conjugate

heat transfer is performed.

Acknowledgement

This project has received funding from

the Euratom research and training

programme 2014-2018 under grant

agreement No 654935 and from the

AREVA Nuclear Professional School.

References:

[1] Chen, S.; Todreas, N.; Nguyan, N.

(2014). Evaluation of existing correlations

for the prediction of pressure drop

in wire-wrapped hexagonal array pin

bundles. Nuclear Engineering and

Design 267, pp. 109 – 131

[2] Rehme, K. (1973). Pressure drop

correla tions for fuel element spacers.

Nuclear Technology 17, 15–23.

[3] Baxi, C.B., Dalle Donne, M., (1981).

Helium cooled systems, the gas cooled

fast breeder reactor. In: Fenech, H. (Ed.),

Heat Transfer and Fluid Flow in Nuclear

Systems. Pergamon Press Inc.,

pp. 410–462.

[4] Cheng, S.-K.; Todreas, N. (1986). Hydrodynamic

models and correlations for

bare and wire-wrapped hexagonal rod

bundles - Bundle friction factors,

subchannel friction factors and mixing

parameters. Nuclear Engineering and

Design 92 (2), 227 – 251.

[5] Kirillov, P.L., Bobkov, V.P., Zhukov, A.V.,

Yuriev, Y.S., (2010). Handbook on

Thermo hydraulic Calculations in

Nuclear Engineering. Thermohydraulic

Processes in Nuclear Power Facilities,

vol. 1. Energoatomizdat, Moscow.

[6] Kamide, H.; Hayashi, K.; Toda, S. (1998).

An experimental study of intersubassembly

heat transfer during

natural circulation decay heat removal

in fast breeder reactors. Nuclear

Engineering and Design 183, 97 – 106.

[7] http://sesame-h2020.eu/

[8] Pacio, J, et. al. (2016), Deliverable 2.10 –

KALLA Inter- wrapper flow setup for

SESAME (thermal hydraulics Simulations

and Experiments for the Safety

Assessment of MEtal cooled reactors)

project, activity: NFRP-01-2014

Improved safety design and operation

of fission reactors, H2020 Grant

Agreement Number: 654935.

Authors

Abdalla Batta

Andreas G. Class

AREVA Nuclear Professional School

Karlsruhe Institute of Technology

Karlsruhe, Germany

OPERATION AND NEW BUILD 229

Operation and New Build

Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class

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