Tb

PEI

a

AA

KBDFT

1

crwwrpltdTrbmc[(

0d

Fusion Engineering and Design 84 (2009) 573–577

Contents lists available at ScienceDirect

Fusion Engineering and Design

journa l homepage: www.e lsev ier .com/ locate / fusengdes

hermal–hydraulic and thermo-structural analysis of first wall for Indian DEMOlanket module

aritosh Chaudhuri ∗, Chandan Danani, Vilas Chaudhari, C. Chakrapani, R. Srinivasan, I. Sandeep,. Rajendra Kumar, S.P. Deshpande

nstitute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat, India

r t i c l e i n f o

rticle history:vailable online 20 January 2009

eywords:lanketEMOirst wallhermal–hydraulics

a b s t r a c t

The first wall (FW) is one of the most important components of any fusion blanket design. India hasdeveloped two concepts of breeding blanket for the DEMO reactor: the first one is Lead–Lithium cooledCeramic Breeder (LLCB), and the second one is Helium-Cooled Ceramic Breeder (HCCB) concept. Both theconcept has the same kind of FW structure. Reduced Activation Ferritic Martensitic steel (RAFMS) used asthe structural material and helium (He) gas is used to actively cool the FW structure. Beryllium (Be) layerof 2 mm is coated on the plasma side of the FW as the plasma facing material. Cooling channels runningin radial–toroidal–radial direction in the RAFMS structure are designed to withstand the maximum Hepressure of 8 MPa. Heat transfer coefficients (HTC) obtained form the correlations revealed that requiredcooling could be achieved by artificially roughened surface towards the plasma-side wall of He cooling

channel which helps to keep the RAFMS temperatures below the allowable limit. A 1D analytical and 2Dthermal–hydraulic simulation studies using ANSYS has been performed based on the heat load obtainedfrom neutronics calculations to confirm the heat removal and structural integrity under various conditionsincluding ITER transient events. The required helium flow through the cooling channels are evaluatedand used to optimize the suitable header design. The detail design of FW thermal–hydraulics, thermo-

e flo

structural analyses, and H

. Introduction

One of the key missions of the ITER is to validate the design con-epts of tritium breeding blankets relevant to a power-producingeactor like DEMO. Indian participation in the ITER project asell as participation in Test Blanket Module (TBM) programmeill provide sufficient confidence on the developments of mate-

ial performance, design and associated technologies for its fusionrogram. In parallel, a national level fusion programme has been

aunched to establish synergy among research institutes and indus-ries for carrying out the R&D activities in the critical areas likeesign of tritium breeding blankets relevant to DEMO [1]. IndianBM program in ITER is one of the major steps in Indian fusioneactor programme towards DEMO and power plant vision [2]. Twolanket concepts have been considered for the design, develop-

ent and testing in ITER for DEMO relevancy. The primary blanket

oncept is Lead–Lithium cooled Ceramic Breeder (LLCB) blanket3,4] and the other one is the conventional solid blanket conceptHelium-Cooled Solid Breeder – HCSB) [5] with variance in geo-

∗ Corresponding author. Tel.: +91 79 23962133; fax: +91 79 239622277.E-mail address: paritosh@ipr.res.in (P. Chaudhuri).

920-3796/$ – see front matter © 2008 Elsevier B.V. All rights reserved.oi:10.1016/j.fusengdes.2008.11.093

w distribution network will be presented in this paper.© 2008 Elsevier B.V. All rights reserved.

metrical design. Both the concept has the same kind of first wall(FW) structure. The FW is designed to withstand the energetic par-ticle fluxes and heat fluxes from the plasma, high thermal andmechanical stresses and magnetic forces during plasma disrup-tions.

1D analytical and the 2D finite element (FE) simulation stud-ies using ANSYS [6] have been performed based on the neutronicsheat load and surface heat flux from plasma on the FW struc-ture to confirm the efficient heat removal from blanket module.3D thermo-structural analysis also has been performed to estimatethe thermal stresses and to verify the structural integrity of the FWstructure. The objective of this study is to estimate the the steadystate and transient temperature distribution on the FW structure,and to optimise the helium flow requirement. The details of theoptimization of thermal–hydraulic parameters, coolant flow pathsare presented in this paper.

2. First wall in the blanket module

The FW assembly is designed to withstand the heat flux fromthe plasma and neutronics heat generation on the FW structureto maintain its temperature below the allowable limits. The typ-ical dimensions of one DEMO blanket is ∼1.7 m (poloidal) × 1.0 m

574 P. Chaudhuri et al. / Fusion Engineering and Design 84 (2009) 573–577

Fig. 1. (a) Schematic of DEMO blanket (b) the top-view of a module at outboard mid-plane.

um ci

(usotDp

TM

P

FFTTFHCCCC

of 20 mm × 20 mm cross section, and a Be layer of 2 mm coated

Fig. 2. Schematic of FW heli

toroidal) × 0.5 m (radial). The DEMO blanket consists of two mod-les resembles the two TBM dimensions are placed toroidallyide-by-side as shown in Fig. 1(a). Fig. 1(b) shows the schematic

f top-view of a module placed at outboard mid-plane. Allhermal–hydraulic calculations have been performed based onEMO relevant neutronics heat load and surface heat flux fromlasma on one such DEMO module as shown in Fig. 1(b).

able 1ain parameters of the FW in a DEMO blanket module.

arameters Value

W structural material RAFMSW dimension (poloidal × toroidal × radial) ∼1.7 m × 0.5 m × 0.5 mhickness of FW structure 28 mmhickness of Be coating 2 mmW surface area facing plasma (m2) 0.85eat flux on the FW (MW/m2) 0.5oolant fluid Helium gasooling channel dimension 20 mm × 20 mmoolant inlet pressure 80 bar (8 MPa)oolant inlet/out (◦C) 300/380

rcuit in the blanket module.

The U-shaped FW structure as shown in Fig. 2(a), is composed ofa 28 mm thick RAFMS structure, having internal cooling channels

on the plasma side of the FW. The coolant channels are designedto allow multiple passes of helium coolant across the FW in orderto maximize the heat removal. The number of helium passes has

Fig. 3. Exit temperature of helium for different FW cooling layouts.

P. Chaudhuri et al. / Fusion Engineering and Design 84 (2009) 573–577 575

Table 2Neutronics heat loads and heat flux on the FW structure.

Material Heat loads (MW)

Be 0.027Front wall 0.126Side wall 0.047Heat flux (@ 0.5 MW/m2) 0.402Total 0.602

brflyM

3

TcFitmacaHeiFF8

c(p

een optimized such that the maximum temperature in the RAFMSemains below the design limit of 550 ◦C. Based on the surface heatux and neutronic heat generation on FW, thermal–hydraulic anal-ses have been carried out for the normal and extreme conditions.ain parameters of FW structure are tabulated in Table 1.

. He flow distribution network in FW structure

The design of this FW cooling scheme is adopted from US DCLLest Blanket Module [7]. The FW structure is having 64 heliumoolant channels, which are divided into two circuits as shown inig. 2(b). One circuit (circuit 1) of the He flow channels have open-ngs at the edge face of the FW and other circuit (circuit 2) havehe channel openings on the inner face of the FW (Fig. 2(c)). This

ulti-pass arrangement is to meet the heat transfer requirementsnd maintain a minimum temperature of the FW. These two cir-uits are always in a counter flow arrangement in order to achieveuniform temperature distribution across the FW surface. The twoe circuits flowing through the FW channels are separated fromach other and only mixed in the outlet manifold prior to enter-ng into the outlet pipe. The manifolds are designed to cover theW height with four consecutive passes. The cooling channel in theW module is designed to withstand the maximum He pressure of

MPa.

Each circuit is having four passes and each pass contains eightooling channels. In circuit 1 the He starts from the bottompoloidal) of the FW and flows from one pass to another pass (sayass one to pass two). This continues four times until the He reaches

Fig. 4. Temperature profile on F

the top location (poloidal) of the last pass as shown in Fig. 2(c). Simi-larly the flow in circuit 2 starts with the pass one and continues fourtimes until it reaches the bottom location (poloidal) of the last passin this circuit as shown in this same figure (Fig. 2(b)). The flow of Heis in opposite direction in two different cooling circuits. Differentflow parameters and various cooling layouts have been examined toselect the optimum thermal–hydraulic parameters and tube layoutfor FW cooling. Fig. 3 shows the He exit temperature and mass flowrequired for two different cooling layouts (four passes and eightpasses).

4. Steady state thermal–hydraulics of FW in blanket module

The thermal–hydraulic system is designed to remove all the heatdeposited in the FW structure in blanket module. The heat loads onFW comprises of (i) surface heat flux and (ii) the nuclear heating onthe front wall and side wall structure. There the total heat loads onDEMO FW structure is 0.602 MW (as shown in Table 2) comparedto the total of 2.24 MW deposited in the DEMO module. Most of theheat on the FW is due to the surface heat flux from plasma only.

1D analytical analysis has been performed to evaluate thethermal performance of FW, specifically, the maximum FW temper-ature, helium outlet temperature and the Heat Transfer Coefficient

(HTC) in the channels. Calculation starts with a specified veloc-ity, inlet coolant temperature and other coolant properties. Thenthe mass flow rate is computed by the equation, m = �VAx, whereAx is the cross-sectional area of the cooling channel. In order todetermine the steady state exit temperature, the following energy

W for DEMO heat fluxes.

576 P. Chaudhuri et al. / Fusion Engineering

b

Q

cahpc

Q

w

N

tcHpnhDa

Fig. 5. Helium flow pressure drop in one single channel of FW structure.

alance equation is used [8]:

= mHeCp.He(THe.out − THe.in)

The helium flow in each of these cooling channels in the FWooling circuit is optimized as ∼0.1 kg/s corresponding to an aver-ge flow velocity of 45 m/s. Therefore, the total mass flow rate ofelium in each pass in one circuit is ∼0.81 kg/s. The fluid tem-erature directly influences the temperature distribution in theomponents through Newton’s law of cooling:

= hAs(tw − tb).

here As is the channel surface area.

u = hd

k= (1/8)�(Re − 1000)Pr

1 + 12.7√

(1/8)�(Pr2/3 − 1)

In order to correctly evaluate the temperature distribution inhe blanket, an accurate estimation of the convective heat transferoefficient between the structures and the coolant is required. TheTC correlations are usually expressed in terms of dimensionless

arameters, that is the Nusselt (Nu), Reynolds (Re) and Prandtl (Pr)umbers. The Pr of helium for our FW cooling circuit is 0.66 andence, Gnielinski correlation has been used (instead of well-knownittus–Boeltus correlation) whose range of validity is 0.5 < Pr < 1.5nd 2300 < Re < 106 [9].

Fig. 6. FE discretization and the constraints of

and Design 84 (2009) 573–577

Where h = HTC, d = diameter, k = thermal conductivity and � isthe drag coefficient given by:

� = 1

(1.82 log Re − 1.64)2

It should be noted that the RAFMS temperature estimated byusing HTC (3160 W/m2 K) obtained for above correlations exceedsits allowable temperature value, so, it is necessary to increase theHTC to keep the RAFMS within its limiting value. To enhance the HTCand to keep the maximum RAFMS temperature below the allowablelimit, the inside surface of the He channels are artificially rough-ened. The Computational Fluid Dynamics (CFD) software FLUENT[10] was used to estimate the HTC on the plasma-side channelwalls having roughness constant of 0.25 mm (same as uniformsand-grain roughness height of 0.25 mm) while all other channelwalls were smooth. It should be noted that the surface roughnessof the cooling channel strongly affects the HTC. Heat transfer in theFW is increased by more than 125% by roughening the surface ofthe plasma-side helium flow channels. Roughening of course, alsoincreases the friction factor and thus the power requirement forthe pump. The HTC obtain by CFD analysis (using FLUENT) withrespect to the fluid bulk temperature is 7119 and 3160 W/m2 K forroughened and smooth surface respectively. HTC values obtained byanalytical analysis (using Gnielinskii correlation) are 3155 W/m2 Kwhich is in good agreement with the value obtained by CFD analysis.

In this analysis, the thermal conductivity of Be and RAFMSwere considered as 110 and 29 W/m K respectively. HTC obtainedfrom the correlations was applied on the cooling channel surfaceand the maximum temperature on different material was esti-mated. Fig. 4 shows the maximum temperature of Be and RAFMSfor smooth channel (without roughening the plasma-side surface)600 and 590 ◦C respectively. It reveals that, to keep the maximumRAFMS temperatures below the allowable limit some heat trans-fer enhancement technique is required. Roughening of the wall isone of the proven heat transfer enhancement technique, which isadopted in FW cooling in the blanket design. To simulate this HTCof ∼7119 W/m2 K (obtained from CFD analysis) is applied on theplasma-side surface of the channels; while all other three smoothchannel walls were estimated as ∼3160 W/m2 K. The maximumtemperature of Be and RAFMS after being artificially roughened theplasma-side surfaces are 521 and 512 ◦C respectively which is well

below the allowable temperature limit of RAFMS. The comparisonof temperature profile obtained on the FW structure for smooth androughen tube is also shown in Fig. 4. Flow analysis is carried out forhelium flow channel in one single FW channel with velocity inletof 45 m/s. The results of total gauge pressure contours are shown in

the FW used for thermal stress analysis.

P. Chaudhuri et al. / Fusion Engineering and Design 84 (2009) 573–577 577

FW st

Ft∼t0b

5

kFsv(yof3∼plp

6

a

nafi

Fig. 7. Von Misses Stress in the

ig. 5. In CFD analysis velocity is given as the inlet boundary condi-ions and the pressure drop is calculated. It gives pressure drop of10,000 Pa in a single channel (as shown in Fig. 5) due to friction and

wo 90◦ bends. The total pressure losses calculated analytically are.0099 MPa which is in good agreement with the results obtainedy CFD (using FLUENT).

. Thermal stress analysis in FW of blanket module

Thermal structural analysis has been carried on the FW of blan-et module using ANSYS. Fig. 6 shows the FE discretization of theW model and the constraints (all DOF, Ux, Uy, Uz) used for thermaltress analysis. The temperature dependent material properties,iz., specific heat at constant pressure (Cp), thermal conductivityk), thermal expansion coefficient (˛) and Young’s modulus (E),ield strength (Sy) are considered for analysis. The temperature databtained from above thermal–hydraulic analysis is used as the inputor the thermal stress analysis. The maximum stress in the FW is50 MPa at the fillet region, as compared with the yield strength is400 MPa at 510 ◦C as shown in Fig. 7. This stress value is below theermissible limits for the requirements of structure strength regu-

ations according to the 3Sm rules of ASME code for the boiler andressure vessel [11].

. Conclusions

The FW thermal–hydraulic analyses carried out using the 1D

nalytical and 2D FE modeling have been presented.

It is observed that with the smooth surface of the cooling chan-el HTC is very less. So, to keep the RAFMS temperature within itsllowable limits, the surface of the ‘plasma-side’ channel wall is arti-cially roughened by 0.25 mm. This preliminary analysis gives very

[[

ructure in the blanket module.

much useful information in order to optimize the blanket FW designfrom thermal mechanical point of view. The optimization study forefficient heat extraction from FW structure has been carried out byvarying the flow network and coolant flow parameters and opti-mum thermal and flow profiles for the given thermal inputs andmaterial constraints have been obtained. In addition to the steadystate analysis, a transient event with a localized elevated heat fluxon the FW has to be analyzed in details, which will be responsi-ble for the temperature excursions of the different components inFW structure. The accident analysis of helium leakage to the plasmawill be estimated as the future work. A number of critical issues alsohave been identified and they will be investigated in the future.

References

[1] Design Description Document for ‘Indian Lead–Lithium cooled Ceramic Breeder(LLCB) Blanket’, version–1.0, Report to the ITER Test Blanket Working Group(TBWG), April 2008.

[2] R. Srinivasan, S.P. Deshpande and the Indian DEMO team, Strategy for the IndianDEMO design, Fusion Eng.Des. 83 (2008) 889–892.

[3] E. Rajendra Kumar, C. Danani, I. Sandeep, Ch. Chakrapani, N. Ravi Pragash, V.Chaudhari, et al., Preliminary design of Indian Test Blanket Module for ITER,Fusion Eng. Des. 83 (2008) 1769–1772.

[4] C.P.C. Wong, J.-F. Salavy, Y. Kim, I. Kirillov, E. Rajendra Kumar, N.B. Morley, etal., Overview of liquid metal TBM concepts and programs, Fusion Eng. Des. 83(2008) 850–857.

[5] P. Chaudhuri, Status of Engineering Design & Analysis for Indian TBMs, Pre-sented at ITER TBWG-20 Meeting, November 5–7, 2008, Aix-en-Provence,France.

[6] ANSYS User’s Manual, ANSYS Inc.[7] US Design Description Document: Dual Coolant Pb-17Li (DCLL) TBM, November

15, 2005.[8] M.N. Ozisic, Heat Transfer: A Basic Approach, McGraw Hill Book Co., 1985.[9] G. Aiello, F. Gabriel, L. Giancarli, G. Rampal, J.-F. Salavy, Thermal–hydraulic anal-

ysis of the HCLL DEMO blanket, Fusion Eng. Des. 82 (2007) 2189–2194.10] Fluent, Users guide 6.3, Fluent Inc., 2006.11] Am. Soc. Mech. Eng., ASME, Boiler and Pressure Vessel Code, Section III, 2004.