WWR-M2 Nuclear Research Reactor Power Upgrading - Thermal Hy

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WWR -M2 Nuc lear Resear ch Reacto rPower Upg rad ing - T he rma l Hydr aul ic

Anal ys isF. M. BSEBSU , and G . BEDE

Department for Energy

Budapest University of Technology and Economic

H-1521 Budapest, Hungary

Fax: +36 1 463-3273, Phone +36 1 463 2594

Email:[email protected]

Abstract

This paper presents the outline of the core thermal hydraulics design and

analysis of Budapest nuclear research reactor (WWR-M type), which is a tank type,

light water, cooled 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 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 UAlx-Al fuels in WWR-SM fuel coolant channel the possibility of force up the

reactor power to 20 MWth was investigated with keeping the same core configurationand with three types of fuel coolant channel. The study was carried out for an

equilibrium core, with compact load (223 fuel assemblies) under normal operation

conditions.

Keywords:Research reactor, reactor thermal hydraulics, reactor heat transfer, and

reactor power upgrading, WWR-M2 channel.

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mailto:[email protected]:[email protected]


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1. INTRODUCTION

In this report we shall present the theoretical outline of the core power

uprating thermal hydraulics design and analysis of WWR-M2 research reactor as

shown in Figure 1, [1-5] 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 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 uprating too.[1,2]

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

UAlx-Al fuels in WWR-SM fuel coolant channel (as shown in Figure 2) the possibility

of force up the reactor power to 20 MWth was investigated with keeping the same core

configuration and with new design type of fuel coolant channel.

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

assemblies) under normal operation conditions. 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 MW th 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 1992 1993. The design concept of the new

reactor (upgrade one) is that it has great, flexibility of utilization and that it provides

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an adequate neutron flux for isotope production, material testing, and neutron physics

measurement. The performance of upgraded reactor has been investigated using the

WWR-SM fuel type as given in Table 1, with 10 to 20 MWth power level.

2. THERMAL HYDRAULICS CALCULATIONS

Thermal hydraulic studies for the steady state condition were made using the

THMOD2 code [6-8] developed for thermal hydraulic analysis of nuclear research

reactor with fuel coolant channel type WWR-SM under low temperature and low

pressure coolant conditions. 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:

V TCP p = 1

Where:

v = Total coolant volume flow rate, [m3/hr]

P = Reactor core thermal power, [kW]Cp = Average specific heat of coolant = 4.19 [Kj/Kg.

C]

= Average coolant density = 988 [Kg/m3]

T = Temperature difference = Tout - Tin [C].

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:[3-5]

1)2bzf(

2V-)()(

2

++++= endo zWgPzP 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 and z is the channel axial distance.

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The inlet pressure is 1.512 [bar], and reactor core parameters are shown in Table 2,

and coolant velocity is calculated by THMOD2 code for each volume flow rate and

the reactor core configuration.

The Dittus-Boelters, and X59 [8,9] correlationswere used for the calculation of

the convection heat transfer coefficient. The BerglesRohsenows, and H95

correlations [8,9] for the Onset of Nucleate Boiling (ONB) temperature, and the existing

international and X-2000 [11] correlations for DNB heat flux calculation. Boiling

temperature, and saturation temperature. i.e. the complete reactor core heat transfer

package modelling is described in THMOD2 code. [12]

3. 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 1 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 a function of the

reactor core power according to Eq. 1.

Starting at 10 MWth the 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 1 as a sample problems of THMOD2 code for thermal

hydraulics analysis of WWR-M2 research reactor core and from these we shall select

the optimal fuel element dimensions are 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 may have

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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. [13]

Table 3 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 example for illustration, and from this

table we shall consider only three types of fuel coolant channels for reactor power

uprating thermal hydraulics analysis according to reactor core design operating limits.

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 4. The consequence of these results

in Table 4 we shall consider the WWR-SM1 fuel coolant channel for WWR-M2

research reactor core power uprating thermal hydraulics analysis.

The fuel cladding surface temperature, saturation temperature and ONB

temperature as a function of reactor power level and reactor coolant inlet temperature

by using the WWR-SM1 fuel coolant channel type. This channel type is good and

suitable for WWR-M2 research reactor normal and uprating operation conditions

according to reactor core design 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 as (P = 14 MWth, and Tin = 40C).

4. WWR-M2 REACTOR UPRATING THERMAL HYDRAULICS STUDIES

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

reactor core of WWR-M2 at 10 MW th with 36 % low enrichment uranium (Russian

standard) fuel to investigate the thermal hydraulic analysis and reactor core

performance. The WWR-M2 research reactor is a pool-type, light water-cooled and

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moderated nuclear research reactor core. The heat transfer correlations, ONB

temperature correlation and DNB critical heat flux correlations adopted in the

THMOD2 code for thermal hydraulic analysis and design of nuclear research reactor.

Thermal hydraulics calculations were carried out for the forced-convection cooling

mode with steady state conditions at core power of 10 MWth with fuel coolant channel

WWR-SM1.

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.

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. 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 are little in magnitude and only Tsat becomes lower according to the

pressure decrease due to the increase of coolant velocity. Therefore, both of T sat 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 TONB is higher the fuel surface temperature at the coolant

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velocity of 4.5 9 m/sec. In this range of coolant velocity, no boiling occurs in the

sub-channel and 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-SM1 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

bulk temperature along the WWR-SM1 fuel coolant channel with the operating

coolant velocity as shown in Figure 6. The relationship of Nu vs. Re and heat transfer

coefficient 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 MW th, Tin = 40C. Figure 8

illustration the various DNB heat flux correlations described in the heat transfer

package of the THMOD2 code for the condition of the thermal hydraulic analysis of

WWR-M2 nuclear research reactor fuel coolant channel.D. As for the core exist

temperature of coolant, one should be careful of the following problem. If the coolant

temperature is considerably high at 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 bulk 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 MW th. In the condition of normal operation with

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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. 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.

Core thermal hydraulic characteristics thus designed and analyzed for the

forced-convection cooling mode at the reactor core power level of 14 MW th are

summarized in Table 5.

5. CALCULATION RESULTS

On the bases of the results obtained using THMOD2 code calculation for WWR-M2

Nuclear Research Reactor core power uprating. We can conclude that theoretically it

is possible to increase the reactor core 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) and new design fuel coolant channel as given in

Table 1. Also according to the opinion of Russian experts about the uprating of tank

pool reactor type it is possible to increase the reactor power by using existing type of

fuel coolant channel (old design), and by using other new design type of fuel

assemblies.

Finally, we can conclude that the reactor upgrading (WWR-M Reactor) is suitable for

experimental work but from some safety points of view, we have to consider the

following:

1. Increasing the reactor power level or changing the fuel type, the reactor will

become new, and it will not be old reactor.

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Organized by the European Nuclear Society, Binges, Belgium, Nov. 29-31

(1998).

3. Eryekalov, A. N., et. al.: Thin - Walled Fuel Elements WWR-M5 for Research

Reactors, Atomic Energy. Vol. (60) 2, pp. 103 - 106, (1986), (In Russian),

4. Verkhovyekh, P. M. et. al.: Remarks to the Reconstruction of Active Zone in

the Nuclear Reactor Type WWR-M, Atomic Energy, Vol. 41,(3), pp. 201-203,

(1976), (In Russian).

5. A. A. Enin, and et al.: Design and experience of HEU and LEU fuel for

WWR-M reactors, Nucl. Eng. and Des., Vol. 182, pp. 233-240, (1998).

6. BSEBSU, F. M., and BEDE, G.: A Simple Computer Program for the

Calculations of Reactor Channel Temperature Distribution, Periodic

Polytechnica Series Mech. Eng. (1997) Vol. (41), No. 2, pp.133-142, Technical

University of Budapest, Hungary.

7. BSEBSU F. M., and BEDE, G.: Nuclear Reactor Channel Modelling Using

THMOD2 code, KERNTECHNIK, 64, (1999).

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

Research Reactors, Ph. D. Dissertation, Budapest University of Technology

and Economics, Budapest-Hungary, (2001).

9. 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., Technical University of

Budapest, Hungary. Under Press, (2001).

10.BSEBSU, F. M., RAMADAN M. M., and BEDE G.: Tajoura Reactor Power Up

ratingThermal Hydraulic Analysis, International Multidisciplinary

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Conference on Environmental and Economical Development in Libya and

Hungary, Godollo, Hungary, April 27-28, (1998).

11.BSEBSU, F. M., and BEDE, G:, Critical Heat Flux Correlations for Nuclear

Research Reactors, under preparation, (2000).

12. BSEBSU, F. M.: THMOD2 Code Operation Manual, Internal Report,

Department for Energy, Technical University of Budapest, Budapest, Hungary.

(1998).

13. Y. Sudo, H. Ando, H. Ikawa, and N. Ohnishi:, Core Thermohydraulic Design

with 20% LEU Fuel for Upgraded Research Reactor JRR-3, Journal of Nucl.

Sci. and Tech., 22 (7), pp. 551-564, (July 1985).

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

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Fig. 2. WWR-SM Fuel Coolant Channel

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Table 1 WWR-SM fuel coolant channels, fuel meat and clad dimensions [mm]

CHANNEL FUEL E. I FUEL E. II FUEL E. III

TYPE CTH FTH CTH CTH FTH CTH CTH FTH CTH

WWR-SM0 0.90 0.70 0.90 0.90 0.70 0.90 0.94 0.74 094

WWR-SM1 1.092 1.098 0.956 1.092 1.098 0.956 1.092 1.098 0.956

WWR-SM2 1.026 0.847 1.026 1.026 0.847 1.026 1.026 0.847 1.026

WWR-SM3 0.90 0.70 0.90 0.90 0.70 0.90 0.90 0.70 0.90

WWR-SM4 0.80 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

Table 1. Core Design Description Parameters

Reactor type Tank type

Power level, MW 10

Vertical positions 397

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Fuel positions 223

Irradiation position 51

Beryllium displacers 123

Horizontal beam 10

Radial 8

Tangential 2Fuel

Type WWR-SM

Meat Material UAlx-Al

Clad Material Al (SAV-I)

Active Length, mm 600

Lattice Pitch, mm 35

Moderator, coolant H2O

Reflector Beryllium

Control Rod Absorber B4C (18)

Safety Rod 3Automatic Rod 1

Manual Rod 14

Coolant inlet Temperature. C 35

Coolant inlet Pressure, bar 1.52

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Fig. 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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Table 3 The comparison between the centerline temperatures, fuel cladding surface

temperatures, saturation temperatures, ONB temperatures, and boiling temperatures

at P = 10 MWth, and Tin = 50Cfor thefuel coolant channels.

channel type P, [MWth] TF, [C] TCl, [C] Tsat, [C] TONB, [C] TBLG, [C]WWR-SM0 10 155.03 109.22 109.04 111.39 138.80WWR-SM1 10 139.59 102.30 106.60 109.01 138.58WWR-SM2 10 144.50 103.91 107.94 110.31 136.79WWR-SM3 10 153.62 108.24 109.20 111.52 139.04WWR-SM4 10 154.84 109.46 109.20 111.52 139.04WWR-M51 10 184.08 123.03 110.83 113.17 142.01WWR-M52 10 190.85 125.61 111.10 113.40 142.54WWR-M53 10 191.13 125.88 111.10 113.40 142.54

Table 4The fuel centerline temperature and fuel cladding surface temperature as a

function of reactor core power level, coolant inlet temperature and fuel coolant

channel type.

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Fuel P Tin =35C Tin =40

C Tin =50C Tin =35

C Tin =40C Tin =50

C

Type MWth Fuel 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-SM0 15 164.72 164.77 165.89 108.11 110.61 116.21

18 170.78 170.64 171.41 112.04 114.44 119.81

20 174.51 174.26 174.82 114.50 116.85 122.0810 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-SM1 15 146.61 147.51 150.16 100.49 103.41 109.66

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-SM2 15 152.21 152.76 154.77 102.00 104.75 110.68

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 5. Summary of core thermal hydraulics analysis at 14 MWth for WWR-M2

research reactor core

Primary system total volume flow rate, [m3/hr] 2359Flow ratio in active core region, [%] 78

Coolant velocity in WWR-SM1 sub-channels, [m/sec] 4.75

Core inlet coolant temperature, [oC] 40

Average temperature through primary circuit system, [oC] 5

Core inlet pressure, [bar] 1.512

Pressure loss through active reactor core, [bar] 0.232

Minimum temperature margin to ONB, [oC] 5

Minimum DNB ratio, [--] 1.86

Maximum cladding surface temperature (upper limit), [oC] 104

Core exit coolant temperature, [o

C] 55.55Onset Nucleate Boiling temperature, TONB, [

oC] 109

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Saturation temperature, Tsat, [oC] 104

ONBq , [W/cm2] 108.8

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

velocity of sub-channel C.

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Fig. 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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Fig. 7. Illustration of heat transfer correlation applied for forced-convection single-

phase flow for down flow.

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Fig. 8. Illustration of DNB critical heat flux correlation used for sub-channel D of

WWR-SM1 fuel coolant channel.

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Fig. 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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Fig. 10. Calculated results of maximum cladding surfaces of the fuel element 3 of

WWR-SM1 fuel coolant channel vs. core coolant velocity.

WWR-M2 Nuclear Research Reactor Power Upgrading - Thermal Hy

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