Thermal Hydraulic Analysis and Design of WWR-M2 Nuclear Research Reactor - Power Uprating

8/14/2019 Thermal Hydraulic Analysis and Design of WWR-M2 Nuclear Research Reactor - Power Uprating

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Thermal Hydraulic Analysis and Design ofNuclear Research Reactor - Power Up-rating PROCEDURES

BSEBSU, Farag Muftah

Reactor Department, Tajoura Research Center

P.O. Box 30878, Tajoura (Tripoli) Libya

Fax: +218 21 360-4141, Phone +218 21 360 4142

Email:[email protected]

Abstract

This report presents the outline of the core thermal hydraulic

design and analysis (Operational Safety Analysis) of Budapest

nuclear research reactor (WWR-M2 type), which is a tank type, light

water-cooled nuclear research reactor with 36% enriched uranium

coaxial annuli fuel. The Budapest nuclear research reactor is

currently upgraded to 10 MWth of thermal power, while the cooling

capacity of the reactor was designed and constructed for 20 MWth.

This reserve in the cooling capacity serves redundancy today but can

be used for future upgrading too. The core thermal hydraulic design

was, therefore, done for the normal operation conditions so that fuel

elements may have enough safety margins both against the onset ofnucleate boiling (ONB) not to allow the nucleate boiling anywhere in

the reactor core and against the departure from nucleate boiling

(DNB). Thermal hydraulic performance was studied, and it is shown

that the 36% enriched UAlx-Al fuels in WWR-SM fuel coolant channel

1
mailto:[email protected]:[email protected]


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does not make possible to force up the reactor power to 20 MWth. The

study was carried out for an equilibrium core, with compact load (223

fuel assemblies) under normal operation conditions only (steady state

condition).

1. INTRODUCTION

In this report we shall present the theoretical outline of the

core power uprating thermal hydraulic design and analysis of WWR-

M2 research reactor,which it is a tank type, light water, cooled reactor

with 36% enriched uranium coaxial annuli fuel. The WWR-M2

nuclear research reactor is currently uprated to 10 MW th of thermal

power, while the cooling capacity of the reactor was designed and

constructed for 20 MWth. This reserve in the cooling capacity serves

redundancy today but can be used for future uprating too.[1]

The reactor was first put into operation in 1959; its principal

functions at that time were to serve as a facility for basic research

experiments in the frameworks of research programs of the Academy

of Science and industrial development projects. The reactor was first

upgraded in 1967, a new type of fuel was introduced and beryllium

reflector was applied, that allowed to increase the reactor thermal

power from 2 MWth to 5 MWth, and after 27 years of operation a full-

scale reconstruction and upgrading project was started. The

reconstructed reactor was re-operated in 19921993. The design

concept of the new reactor (upgrade one) is that it has great, flexibility

of utilization and that it provides an adequate neutron flux for isotope

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production, material testing, and neutron physics measurement. The

performance of upgraded reactor has been investigated using the

WWR-SM fuel type, with 10 to 20 MWth power level. [2]

The core thermal hydraulic design was, therefore, done for the

normal operation conditions so that fuel elements may have enough

safety margins both against the Onset of Nucleate Boiling (ONB) not

to allow the nucleate boiling anywhere in the reactor core and against

the Departure from Nucleate Boiling (DNB). Thermal hydraulic

performance was studied, and it is shown that the 36% enriched UAl x-

Al fuels in WWR-SM fuel coolant channel, dose not make possible to

force up the reactor thermal power to 20 MW th.

The study was carried out for an equilibrium core, with

compact load (223 fuel assemblies) under normal operation

conditions.

2. WWR-M REACTOR CORE OPERATION EXPERIENCE

AND DESCRIPTION

2.1 WWR-M REACTOR CORE OPERATION EXPERIENCE

The pool-type WWR-M reactors serve a wide range of

scientific research and engineering purpose requiring a high neutron

flux. The first reactor type was put in the operation on December

1959, in Leningrad, Russia as described in the International Atomic

Energy Agency (IAEA) documents, (IAEA Research Reactor

WWR-M Leningrad Aug. 1960). The WWR-M reactor has various

fuel coolant channels. The fuel assemblies for this reactor were first

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produced in the late 1950; is initially fuel elements were composed of

Al+UO2, cermets (1959-1963) but from beginning of 1963 the form is

an Al+U alloy (WWR-M2). Their specific heat transfer surface was

almost four times as large as the EK-10 rod elements used earlier in

WWR-S pool reactors. In consequence, the designed power has been

raised from 2 to 10 MWth. From 1967 together with WWR-M2 fuel

elements, new fuel elements WWR-SM were in production with

increased of active length (50 cm to 60 cm), which are still in use in

reconstructed foreign research reactors. Many years of experience

with these and other similar fuel elements in pool research reactors are

availble (WWR-M in Gatchina and Kiev; WWR-Ts in Obninsk;

WWR-K in Alma-Ata; Eva in Poland and WWR-SM in Germany and

Hungary).

The elaboration of new fuel elements type WWR-M was done

in two stages. In the first stage, the optimization of the geometrical

parameters of fuel elements was done leading to an increase of

specific heat transfer surface by factor of 1.8. Some sets of these type

assemblies (WWR-M3) were in service at the WWR-M reactor of the

Petersburg Nuclear Institute of Physics (PNPI) BP Konstantinov in

Gatchina between 1973 and 1980. In the second stage, the design

concentrations of235U in the core was optimized (at 125 g/l), leading

to a large increase in the power production of each assembly (WWR-

M5) and to raise the spare reactivity of the reactor. Table 1 shows the

characteristic of the WWR-M fuel assemblies, and Table 2 shows the

performance of WWR-M fuel assemblies in the WWR-M reactor of

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PNPI. The base of the reactor core is a hexagonal grid plate with 397

identically formed holes. Fuel assemblies and beryllium displacers can

be put into these holes, guide tubes of the control rods as well. The

lattice pitch of 35 mm; the core positions are occupied by the fuel

assemblies, control rods, beryllium displacers and isotope production

channels. Stationary beryllium reflector of 20-cm average thickness

surrounds the core. The cooling water is flowing down stream across

the reactor core. The fuel of the reactor is of the WWR-M type

(Russian product). It is an alloy of aluminum and uranium-aluminum

eutectic with aluminum cladding. The uranium enrichment is 21%,

36%, and 80 %. The first fuel assembly contains two fuel tubes, with

the outer tube (un fueled one) is of hexagonal shape, while the two

inner ones are cylindrical. The reactor core horizontal cross sections is

[10-13]shown in Figure 1. At the end of 1950 years the first Russian

heat exchange assemblies TBC were developed, built from seamless

tubing fuel elements WWR-M1, which were homogeneously filling up

the active zone without occupying its volume for elements, only for

constructional designations. The WWR-SM fuel coolant channel is

three coaxial annuli (fuel elements) and the outer fuel element is a

hexagonal shape with pitching is a 35 mm as shown in Figure 2. [3-

10]

2.2 WWR-M REACTOR CORE DESCRIPTION

The WWR-M is a cylindrical tank type reactor. The reactor

core is placed 5.145 m below the surface of the reactor tank (in order

to minimize the radiative exposure to the personel), which is open to

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atmospheric pressure. The diameter of the tank is 2300 mm, and its

height is 5685 mm. The heavy concrete reactor-shielding block is

situated in a rectangular semi-hermetically sealed reactor hall. The

base of the reactor core is a hexagonal grid plate as shown in Figure 1,

with 397 identically formed holes. The fuel assemblies and the

beryllium displacers can be put into these holes, as well as the guide

tubes of the 18-absorber rods. The equilibrium core size (in this study)

consists of 223 fuel assemblies, and the control rods, beryllium

displacers and isotopes production channels, occupy the remaining

core positions. A fixed beryllium reflector of 20-cm average thickness

surrounds the core. The fuel assembly type is WWR-SM as shown in

Figure 2 (consists of 3 coaxial fuel elements). The innermost is a tube,

this is followed by a second fuel element with an annulus cross-

section, and the third fuel element (outer) is a hexagonal shape. [1-10]

3. THMOD2 COMPUTER CODE DESCRIPTION

The THMOD2 (Thermal Hydraulic Modeling version 2)

[3,11-13] code is a one-dimensional computer program (axial

direction) and it provides a capability for analysis of the steady state

thermal hydraulic analysis of research reactors in which coaxial annuli

and/or plates type fuel elements are adopted. In this code, subroutines

to calculate temperature distribution in fuel elements. The THMOD2

code can calculate fuel temperatures under forced convection cooling

mode with downflow direction. A heat transfer package is used for

calculating heat transfer coefficient, DNB heat flux etc.

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The heat transfer package was especially developed for research

reactors, which operated under low pressure and low temperature

conditions using coaxial annuli and/or plate-type fuel elements, just

like the WWR-M2 reactor.

4. THERMAL HYDRAULIC CALCULATIONS

Thermal hydraulic studies for the steady state conditions were

made using the THMOD2 code with fuel coolant channel type WWR-

SM under low temperature and low pressure coolant conditions (in

this study we consider only the WWR-M2 reactor operatinal safety

analysis, i.e. the possibility of reactor power uprating using the same

and/or exicting fuel assembly at normal operation of the reactor, and

boiling occures or not. The cooling water flows downward through the

reactor core, with inlet coolant temperature of 25-50 C, while the

temperature difference between the core inlet and outlet is a round 5

C with a volume flow rate calculated according to the following

equation:

TcVP p =

(1)

Where V

is the reactor core total coolant volume flow rate, [m3/hr], P

is the reactor core thermal power, [kW], cpis the average specific heat

of coolant = 4.19 [kj/kg.

C], is the average reactor core coolantdensity = 988 [kg/m3], and T is the temperature difference betweenreactor core outlet and inlet = Tout - Tin [

C].

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The dependence Tsat (z) has been calculated by the well-known

dependence of saturation temperature by the pressure depending on

coordinate Tsat [P (z)] where: [7,10]

1)2b

zf(

2

V-)zW(gP)z(P en

2

do ++

++=(2)

where g, is the gravity acceleration, Po is the atmospheric pressure, Wd

is the reactors pool depth, en is the channel entrance friction

coefficient, is the friction factor, b is the water spacing between fuel

plates, V is the reactor core coolant velocity, and z is the channel axial

distance.

The reactor core inlet pressure is 1.512 [bar], and reactor core

parameters are shown in Table 3, and coolant velocity is calculated by

THMOD2 code for each volume flow rate and the reactor core

configuration.

The X59 [3, 14], and Dittus-Boelters [15] correlations were

used for the calculation of the convection heat transfer coefficient. The

H95 [3], and BerglesRohsenows [16] correlations for the Onset of

Nucleate Boiling (ONB) temperature, and the X2000 [3] and existing

international [17-21] correlations for DNB heat flux calculation.

Boiling temperature and saturation temperature (i.e. the complete

reactor core heat transfer package modeling) is described in THMOD2

code operation manual. [3, 13]

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5. PROCEDURE OF THE REACTOR CORE UPRATING

The THMOD2 code considers equal pressure drop for all

channels of the reactor core, and calculates the velocity distribution

for fuel coolant channels, using the dimensions of fuel elements as

given in Table 4 for performing the upgrading calculations. The

calculations were preformed with the assumption that the three main

primary pumps are operating at full load with a total flow rates as afunction of the reactor core power according to Equation (1).

Starting at 10 MWth the reactor core power level was gradually

increased in steps of 1 MWth up to 20 MWth power level, and

according to the maximum operating limits of the WWR-M2 research

reactor for a fuel centerline temperature 150 C and the maximum

cladding surface temperature 104 C. Using the old and new fuel

element dimensions as shown in Table 4 as sample problems of

THMOD2 code, we shall select the optimal fuel element dimensions

suitable for WWR-M2 research reactor power uprating and also

according to the reactor core design operating limits for fuel centerline

temperature and fuel cladding surface temperature. Fuel elements

should have enough safety margins both against the onset of nucleate

boiling (ONB) not to allow the nucleate boiling anywhere in the

reactor core and against the departure from nucleate boiling (DNB).

[3, 22, 23]

Table 5 shows the fuel centerline temperatures, fuel cladding

surface temperatures, saturation temperatures, ONB temperatures, and

boiling temperatures as a function of fuel coolant channel type as an

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example for illustration, and from this table we shall consider only

three types of fuel coolant channels for the reactor power uprating

thermal hydraulic analysis.

The maximum fuel centerline temperature and fuel cladding

surface temperature as a function of reactor core power level and

reactor coolant inlet temperature for three types of fuel coolant

channels are shown in Table 6. The consequence of these results in

Table 4 we will consider the WWR-SM1 fuel coolant channel for

WWR-M2 research reactor core power uprating thermal hydraulic

analysis. [3]

The fuel cladding surface temperature, saturation temperature

and ONB temperature as a function of reactor power level and reactor

coolant inlet temperature and according to reactor core design

conditions and operating limits are shown in Figure 3 and from this

figure we shall select the maximum reactor operating power level and

reactor coolant inlet temperature and other operating parameters as (P

= 14 MWth, Tin = 40C, TONB = 109

C and, and Tsat =104C).

6. WWR-M2 REACTOR UPRATING THERMAL HYDRAULIC

STUDIES

In this section, we are planning to remodel the existing nuclear

research reactor core of WWR-M2 at 10 MWth with 36 % enrichment

uranium (Russian standard) fuel to investigate the thermal hydraulics

and reactor core performance.

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The temperature is shown as a function of coolant velocity

because the coolant velocity is the only dominant variable to the fuel

surface temperature. Both the ONB temperature and the saturation

temperature become lower with an increase of coolant velocity

because an increase in coolant velocity gives lower local pressure

according to the increase of pressure loss.

Figure 4 shows the calculated results of the fuel surface

temperature, ONB temperature and saturation temperature where the

difference between the ONB temperature and the fuel surface

temperature is a minimum for sub-channel C, and it is equal to 5 C.

The pressure at top and the bottom of WWR-SM1 fuel coolant

channel are shown in Figure 5 with the coolant velocity as a

parameter, to show the characteristics of pressure decrease due to the

increase of coolant velocity. The increase of coolant velocity and

decrease of pressure give lower temperature (TONB-Tsat).

But in this case, the effects of an increase of coolant velocity

and decrease of pressure on the increase of temperature difference

(TONB-Tsat), due to the increase of coolant velocity is little in magnitude

and only Tsat becomes lower according to the pressure decrease due to

the increase of coolant velocity. Therefore, both of Tsat and TONB

become lower with the increase of coolant velocity. On other hand, the

fuel surface temperature becomes lower with an increase of coolant

velocity. It should be noticed in Figure 4 that the T ONB is higher the

fuel surface temperature at the coolant velocity of 4.5 9 m/sec. In

this range of coolant velocity, no boiling occurs in the sub-channel and

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on the other hand, two-phase flow occurs with nucleate boiling at the

velocity less than 4 m/sec. Therefore, 4.75 m/sec should be adopted as

design velocity for the WWR-M2 reactor core fuel coolant channel

and with total volume flow rate of 2359 m3/hr. At the design velocity

of 4.75 m/sec thus determined, the pressure drop between the core

inlet and the bottom of WWR-SM1 fuel coolant channel is about

0.2323 bar as shown in Figure 5. The distribution of fuel centerline

temperature, fuel cladding surface temperature and coolant

temperature along the WWR-SM1 fuel coolant channel with the

operating coolant velocity are shown in Figure 6.

We tried to formulate new heat transfer coefficient correlation

X59, and new critical heat flux correalation also using existing

international experimental data and my correlations consider as better

more limiting operation domain [3,14] as former correlations,

therefore, the relationship of Nu vs. Re and heat transfer coefficient

(The new oneX59 and some of international correlations) applied for

forced-convection single-phase flow in down flow direction, for

WWR-SM1 fuel coolant channel with D = 4.71 5.47 mm with active

length = 60 cm, is illustrated in Figure 7 with reactor core power = 14

MWth, Tin = 40C.

Figure 8 illustration the various DNB heat flux correlations

(The new one X2000 and some of international correlations)

described in the heat transfer package of the THMOD2 code. As for

the core exist temperature of coolant, one should be careful of the

following problem. If the coolant temperature is considerably high at

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exist of the core, there is possibility that the coolant temperature

should become the saturation temperature resulting in the two-phase

flow at the location where the local pressure is the lowest in the

primary cooling line. This situation should be avoided for a stable

steady state operation condition.

Figure 9 shows the calculation results of the average coolant

temperature at exist of the fuel coolant channel and the saturation

temperature where the local pressure is the lowest, as the function of

coolant velocity in the fuel coolant channel. The results are shown for

the core power of 14 MWth. In the condition of normal operation with

the coolant velocity of 4.75 m/sec designed for the WWR-SM1 fuel

coolant channel, the lowest pressure is about 1.167 bar with the

saturation temperature of 95.76 C and the average bulk temperature of

coolant at exist of WWR-SM1 fuel coolant channel is about 55.55 C

as shown in Figure 9 and its consequently no boiling occurs in the

primary cooling piping system. The maximum allowable fuel element

cladding surface temperature is about 104 C as shown in Figure 10.

The statistical comparison between the experimental data, X59

correlation and Dittus-Boelter correlation for calculation of Nu

number as given in Table 7.

Table 8 gives a statistical comparison summary between

X2000 correlation and some of international DNB correlations, and

also, Figure 11 shows the comparison between the X2000 correlation

and some experimental data. [3, 24, 25]

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Core thermal hydraulic characteristics [3] thus designed and

analyzed for the forced-convection cooling mode at the reactor core

power level of 10 and 14 MWth are summarized in Table 9.

7. CALCULATION RESULTS

On the bases of the results obtained using THMOD2 code

thermal hydraulic calculation for WWR-M2 Nuclear Research Reactor

core power uprating we can conclude that theoretically it is possible to

increase the reactor core thermal power level up to 14 MWth safely

and without any operational problems of the reactor using the existing

WWR-SM1 fuel coolant channels (3 coaxial fuel elements).

Acknowledgment

The authors would like to express their appreciation to the

Prof. Dr. L. Rdonyi, head of Department for Energy, Budapest

University of Technology and Economics, Hungary for his continuous

encouragement, valuable suggestions and supporting this work. Also,

their thanks are forwarded to Prof. Dr. Tams Jszay and Prof. Dr.

Tams Kornyi for thier suggetions and discussions.

References

1. KFKI, (Central Research Institute for Physics): The Budapest

Research Reactor Safety Report Analysis, Budapest, Hungary,

(1994).

2. Hargitai, T.: Refueling strategy at the Budapest research

Reactor, 2nd International Topical Meeting on Research Reactor

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fuel management, Organized by the European Nuclear Society

(ENS), Binges, Belgium, Nov. 29-31 (1998).

3. BSEBSU, F. M.: Thermal Hydraulic Analysis of Water-Cooled

Nuclear Research Reactors, PhD. Thesis, Budapest University

of Technology and Economics, Budapest-Hungary, (2001 Oct.

17).

4. Eryekalov, A. N., et. al.: Thin-Walled Fuel Elements WWR-M5

for Research Reactors, Atomic Energy. Vol. 60/2, (1986), pp.

103 - 106, (in Russian),

5. Verkhovyekh, P. M., et. al.: Remarks to the Reconstruction of

Active Zone in the Nuclear Reactor Type WWR-M, Atomic

Energy, Vol. 41/3, (1976), pp. 201-203, (in Russian).

6. Enin, A. A., et al.: Design and experience of HEU and LEU

fuel for WWR-M reactors, Nucl. Eng. and Des., Vol. 182,

(1998), pp. 233-240.

7. Eryekalov, A. N., et. al.: Reduction of the Enrichment in Fuel

Elements for WWR-M Reactors, IAEA-SM-310/113P, (1984),

pp. 710-726.

8. IAEA Research Reactors Documents, WWR-M - Leningrad,

(1962), pp. 165-169.

9. Zakharov, A. S., et. al.: Control Rods of A WWR-M Reactor

Fitted with Finned External Fuel Elements, Atomic Energy,

(1993), 74/1, pp. 88 - 90.

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10.Eryekalov, A. N. and Petrov, Yu. V.: Parameters Characterizing

Reactor for Physical Experiments, Atomic Energy, (1968),

25/1, pp. 82 - 84.

11.BSEBSU, F. M., and BEDE, G.: A Simple Computer Program

for the Calculations of Reactor Channel Temperature

Distribution, Periodic Polytechnica Series Mech. Eng.,

Budapest University of Technology and Economics, Budapest,

Hungary, Vol. 41/2, (1997), pp. 133.

12. BSEBSU, F. M., and BEDE, G.: Nuclear Reactor Channel

Modelling Using THMOD2 code, KERNTECHNIK, Vol.

64/5-6, (1999), pp. 269-273.

13.BSEBSU, F. M.: THMOD2 Code Operation Manual, Internal

Report, Department for Energy, Budapest University of

Technology and Economics, Budapest, Hungary. (1998).

14.BSEBSU, F. M., and BEDE G.: Theoretical study in Single-

Phase Forced-Convection Heat Transfer Characteristics for

Narrow Annuli Fuel Coolant Channels, Periodic Polytechnica

Series Mech. Eng., Budapest University of Technology and

Economics, Budapest-Hungary. Under Press, (2002).

15. Todreas N. E., and Kasimi, M. S.: Thermal Hydraulic

Fundamental, Vol. I, & II Hemisphere Publishing Corporation,

NY, USA. (1990).

16.Fenech H., and Rohsenow W. M.: Heat Transfer, Ch.16, pp.

335-418, The Technology of Nuclear Reactor Safety, Vol. 2.

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Reactor Material and Engineering, edited by T. J. Thompson

and J. G. Beckerly, MIT Press, (1973).

17. Sudo Y., and Kaminaga M.: A New CHF Correlation Scheme

Proposed for Vertical Rectangular Channels Heated from Both

Sides in Nuclear Research Reactors, Transactions of the

ASME, Vol. 115, (May 1993), pp. 426-434.

18. Tong L. S.: Prediction of Departure from Nucleate Boiling for

an Axially Non-Uniform Heat Flux Distribution, Journal of

Nuclear Energy, Vol. 21, (1967), pp.241-248.

19. Mishima K., and Nishihara H.: The Effect of Flow Direction

and Magnitude on CHF for Low Pressure Water in Thin

Rectangular Channels, Nucl. Eng. and Des., Vol. 86, (1985),

pp. 165-181.

20. Mishima K., Nishihara H., and Shibata T.:CHF Correlations

Related to the Core Cooling of a Research Reactor, JAERI

M84073, (1983), pp. 312 320.

21. Koweri Y., et. al.: Experimental Study on DNB Heat Flux

Correlations for JMTR Safety Analysis, Int. MTG. on Reduced

Enrichment for Research Reactor and Test Reactor, New Port

RI, 23 27 Sept. (1990).

22. BSEBSU, F. M., et al.: Tajoura Reactor Power Uprating

Thermal Hydraulic Analysis, International Multidisciplinary

Conference on Environmental and Economical Development

in Libya and Hungary, Godollo, Hungary, April 27-28, (1998).

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23. Sudo Y., Ando H., Ikawa H., and Ohnishi N.: Core

Thermohydraulic Design with 20% LEU Fuel for Upgraded

Research Reactor JRR-3, Journal of Nucl. Sci. and Tech., 22/7,

(July 1985), pp. 551-564.

24. Teyssedou A., et al.: Critical Heat Flux Data in a Vertical Tube

at Low and Medium Pressures, Nucl. Eng. and Des., Vol. 149,

(1994), pp. 185-194.

25. Groeneveld D. C., et. al.: The 1995 look-up Table for Critical

Heat Flux in Tubes, Nucl. Eng. and Des., Vol. 163, (1996), pp.

1-23.

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Figure 1. WWR-M2 research reactor core horizontal cross-section.

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Figure 2. WWR-SM Fuel Coolant Channel

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Figure 3. The maximum cladding surface temperature, saturation

temperature, and ONB temperature as a function of reactor core

power level and reactor coolant inlet temperature for fuel elements of

WWR-SM fuel coolant channel dimensions.

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Figure 4. Maximum cladding surface temperature, saturation

temperature, and ONB temperature as a function of reactor coolant

velocity of sub-channel C.

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Figure 5. The pressure at reactor top and bottom as a function of

reactor coolant velocity of sub-channel C.

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Figure 6. The axial distribution of fuel centerline temperature, fuel

surface temperature, and coolant temperature along the coolant sub-

channel D of WWR-SM1 fuel coolant channel.

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Figure 7. Illustration of heat transfer correlation applied for forced-

convection single-phase flow for down flow.

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1 2 3 4 5 6 7 8 9 10 110

100

200

300

400

500

600

700

800

900

1000

1100

1200

Labuntsov

Mirshak

Bernath

Biasi

Bsebsu, X2000

Tong, W3

WWR-M Sub-channel D

Tin= 40

oC

Pin= 1.512 bar

De= 6 mm

Power= 14 MWth

QDNB

[W/cm

2]

Coolant Velocity [m/sec]

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Figure 8. Illustration of DNB critical heat flux correlation used for

sub-channel D of WWR-SM1 fuel coolant channel.

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Figure 9. Calculated results of average core exit coolant temperature

and saturation temperature at lowest pressure in primary coolaing

line vs. core coolant velocity.

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Figure 10. Calculated results of maximum cladding surfaces of the

fuel element 3 of WWR-SM1 fuel coolant channel vs. core coolant

velocity.

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Figure 11. Comparison of X2000 correlation, CHF data with

predictions of correlations and look-up CHF table (L=1.4 m, P = 4.9

bar)

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Table 1. Characteristics of WWR-M assemblies.

Assembly

Type

235

U

[%

]

Fuel element

Wall (Meat)

Thickness

[mm]

Specific heat

Transfer

Surface

[cm2/cm3]

Compositio

n

Uraniu

m

Density

[g/cm3]

235U

Conc. in

core

[g/l]

WWR-M1 20 2.3(0.9) 3.67 UO2+Al 1.5 50

WWR-M2

WWR-SM36 2.5(0.7) 3.67 U+Al 1.33 61.2

WWR-M590

90

1.25(0.53)

1.25(0.39)

6.6

6.6

U+Al

UO2+Al

0.77

1.2

125

125

WWR-

M2E36 2.5(0.9) 3.67 UO2+Al 2 122

WWR-

M5E

36

21

1.25(0.43)

1.25(0.43)

6.6

6.6

UO2+Al

UO2+Al

2

3

102

83

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Table 2. Performance of WWR-M assemblies on the WWR-M reactor

of PNPI.

Characteristic WWR-M1 WWR-M2 WWR-M3 WWR-M5

Operating period 1959-63 1963-79 1973-80 1980-97

Reactor power [MW] 10 16 18 18

Mean (max.) burnup in unloaded

assemblies [%]47(76) 41(91) 28(73) 29(59)

Number of used single assemblies 184 2765 638 2235

Mean power production per assembly

[MWday/Ass.] 9.7 10 7.7 14.7

Total power production [GW-day] 1.8 28 5 32.8

Table 3. Core Design Description Parameters

Reactor type Tank type

Power level, MW 10

Vertical positions 397

Fuel positions 223

Irradiation position 51

Beryllium displacers 123

Horizontal beam 10

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Radial 8

Tangential 2

Fuel

Type WWR-SM

Meat Material UAlx-Al

Clad Material Al (SAV-I)

Active Length, mm 600

Lattice Pitch, mm 35

Moderator, coolant H2O

Reflector BerylliumControl Rod Absorber B4C (18)

Safety Rod 3

Automatic Rod 1

Manual Rod 14

Coolant inlet Temperature. C 35

Coolant inlet Pressure, bar 1.52

Table 4. WWR-SM fuel coolant channels, fuel meat and clad

dimensions [mm].

CHANNEL Fuel Element I Fuel Element II Fuel Element III

TYPE CTH FTH CTH CTH FTH CTH CTH FTH CTH

WWR-

SM00.90 0.70 0.90 0.90 0.70 0.90 0.94 0.74 094

WWR-

SM1

1.09

2

1.09

8

0.95

6

1.09

2

1.09

8

0.95

6

1.09

2

1.09

80.956

WWR-

SM2

1.02

6

0.84

7

1.02

6

1.02

6

0.84

7

1.02

6

1.02

6

0.84

71.026

WWR-

SM30.90 0.70 0.90 0.90 0.70 0.90 0.90 0.70 0.90

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WWR-

SM40.80 0.90 0.80 0.80 0.90 0.80 0.80 0.90 0.80

WWR-M51 0.36 0.53 0.36 0.36 0.53 0.36 0.36 0.53 0.36

WWR-M52 0.43 0.39 0.43 0.43 0.39 0.43 0.43 0.39 0.43

WWR-M53 0.41 0.43 0.41 0.41 0.43 0.41 0.41 0.43 0.41

CTH = Clad Thickness, and FTH= Fuel meat Thickness

Table 5. The comparison between the centerline temperatures, fuel

cladding surface temperatures, saturation temperatures, ONB

temperatures, and boiling temperatures at P = 10 MWth, and Tin = 50

Cfor thefuel coolant channels.

channel

type

P,

[MWth]

TF,

[C]

TCl,

[C]

Tsat,

[C]

TONB,

[C]

TBLG,

[C]

WWR-SM0 10155.0

3

109.2

2109.04 111.39 138.80

WWR-SM1 10139.5

9

102.3

0106.60 109.01 138.58

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WWR-SM2 10144.5

0

103.9

1107.94 110.31 136.79

WWR-SM3 10153.6

2

108.2

4109.20 111.52 139.04

WWR-SM4 10154.8

4

109.4

6109.20 111.52 139.04

WWR-M51 10184.0

8

123.0

3

110.83 113.17 142.01

WWR-M52 10190.8

5

125.6

1111.10 113.40 142.54

WWR-M53 10191.1

3

125.8

8111.10 113.40 142.54

Table 6. The fuel centerline temperature and fuel cladding surface

temperature as a function of reactor core power level, coolant inlet

temperature and fuel coolant channel type.

Fuel

channel P

Tin =35

C

Tin =40

C

Tin =50

C

Tin =35

C

Tin =40

C

Tin =50

C

Type MWthFuel Centerline Temperature

[C]

Clad Surface Temperature

[C]

10 152.74 153.20 155.03 100.65 103.42 109.22

13 160.27 160.47 161.17 105.29 107.89 113.63

WWR- 15 164.72 164.77 165.89 108.11 110.61 116.21

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SM0

18 170.78 170.64 171.41 112.04 114.44 119.81

20 174.51 174.26 174.82 114.50 116.85 122.08

10 135.10 136.34 139.59 92.63 95.73 102.30

13 142.29 143.31 146.17 97.49 100.47 106.84

WWR-

SM115 146.61 147.51 150.16 100.49 103.41 109.66

18 170.78 170.64 171.41 112.04 114.44 119.81

20 174.51 174.26 174.82 114.50 116.85 122.08

10 139.28 141.84 144.50 93.97 97.63 103.91

13 148.01 148.69 150.93 99.24 102.05 108.11

WWR-

SM215 152.21 152.76 154.77 102.00 104.75 110.68

18 157.99 158.36 160.06 105.89 108.54 114.31

20 161.57 161.83 163.34 108.34 110.93 116.61

Table 7. The statistical comparison between the experimental data,

X59 correlation and Dittus-Boelter correlation for calculation of

Nusselt Number.

Correlation Mean Standard

Deviation ()

S. Error

()

Data

No.

X59 114.37 46.03 7.1 42

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Dittus-

Boelter

128.5

255.43 8.6 42

Experimental121.6

255.76 8.6 42

Table 8. Statistical comparison summary between X2000 correlation

and some of international DNB correlations.

QDNB Mean, Standard Deviation S. Error No.

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[W/cm2] () ()

Labuntso

v227.69 18.75 1.50 156

Mirshak 315.38 31.68 2.54 156

Biasi 224.05 59.15 4.74 156

X2000 260.64 40.56 3.25 156

Table 9. Summary of core thermal hydraulic analysis and design for

WWR-M2 research reactor core

Parameter10

MWth

14

MWthPrimary system total volume flow rate, [m3/hr] 1750 2359

Flow ratio in active core region, [%] 78 78

Coolant velocity in WWR-SM1 sub-channels,

[m/sec]3 4.75

Core inlet coolant temperature, [oC] 50 40

Average temperature through primary circuit

system, [oC]5 5

Core inlet pressure, [bar] 1.512 1.512Pressure loss through active reactor core, [bar] 0.173 0.232

Minimum temperature margin to ONB, [oC] 7 5

Minimum DNB ratio, [--] 2.41 1.86

Maximum cladding surface temperature (upper

limit), [oC]104 104

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Core exit coolant temperature, [oC] 64 55.55

Onset Nucleate Boiling temperature, TONB, [oC] 111 109

Saturation temperature, Tsat, [oC] 108.5 104

ONBq , [W/cm2] 108 108.8