BARC IAEA Training Course/Workshop on Natural Circulation in Water Cooled Nuclear Power Plants, ICTP, Trieste, June 25-29,2007 Examples of Natural Circulation

BARCBARC IAEA Training Course/Workshop on Natural Circulation in Water Cooled Nuclear Power Plants, ICTP, Trieste, June 25-29,2007

Examples of Natural Circulation in PHWR

by

Dilip Saha and P.K.Vijayan

Reactor Engineering DivisionBhabha Atomic Research CentreTrombay, Mumbai 400085, INDIA

IAEA Training Course/Workshop on Natural Circulation in Water Cooled Nuclear Power Plants, ICTP, Trieste, June 25-29,2007

BARCBARCCOURSE ROADMAP

Opening Session

INTRODUCTIONS ADMINISTRATION COURSE ROADMAP

Introduction

GLOBAL NUCLEAR POWER ROLE OF N/C IN ADVANCED

DESIGNS ADVANTAGES AND

CHALLENGES

Local Transport Phenomena & Models

LOCAL MASS, MOMENTUM AND ENERGY TRANSPORT PHENOMENA

PREDICTIVE MODELS & CORRELATIONS

Integral System Phenomena & Models

SYSTEMS MASS, MOMENTUM AND ENERGY TRANSPORT PHENOMENA

N/C STABILITY AND NUMERICAL TECHNIQUES

STABILITY ANALYSIS TOOLS PASSIVE SAFETY SYSTEM DESIGN

Natural Circulation Experiments

INTEGRAL SYSTEMS TESTS SEPARATE EFFECTS TESTS TEST FACILITY SCALING METHODS

Reliability &Advanced Computational Methods

PASSIVE SYSTEM RELIABILITY CFD FOR NATURAL CIRCULATION

FLOWS

Opening Session

INTRODUCTIONS ADMINISTRATION COURSE ROADMAP

Introduction

GLOBAL NUCLEAR POWER ROLE OF N/C IN ADVANCED

DESIGNS ADVANTAGES AND

CHALLENGES

Local Transport Phenomena & Models

LOCAL MASS, MOMENTUM AND ENERGY TRANSPORT PHENOMENA

PREDICTIVE MODELS & CORRELATIONS

Integral System Phenomena & Models

SYSTEMS MASS, MOMENTUM AND ENERGY TRANSPORT PHENOMENA

N/C STABILITY AND NUMERICAL TECHNIQUES

STABILITY ANALYSIS TOOLS PASSIVE SAFETY SYSTEM DESIGN

Natural Circulation Experiments

INTEGRAL SYSTEMS TESTS SEPARATE EFFECTS TESTS TEST FACILITY SCALING METHODS

Reliability &Advanced Computational Methods

PASSIVE SYSTEM RELIABILITY CFD FOR NATURAL CIRCULATION

FLOWS


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Natural circulation can take place in a nuclear reactor under following circumstances

• In PHWR core is horizontal leading to horizontal flow path in core. In this case, it is difficult to predict direction of flow when started from stationary condition

• Different rows of channels are at different elevations leading to difference in driving force

• Hydraulic resistance of flow path is large because of the presence of long inlet and outlet feeder pipes and long fuel channels

Experiments on natural circulation have been conducted in NAPS-1, the Indian Pressurised Heavy Water Reactor (PHWR).Rated power of NAPS-1 is 235 MWe

• As design intent• Under accidental condition• Under planned experimental programme

Natural circulation phenomenon is quite complex in a PHWR because

INTRODUCTION


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PRIMARY HEAT TRANSPORT SYSTEM OF PHWR


BARCBARC PHWR CORE CONFIGURATION


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TEST OBJECTIVES

Phenomenological studies were earlier carried out in a simple figure-of-eight loop.Based on the results of these studies a computer code TINFLO –III was developed

Tests were carried out in Unit-1 of Narora Atomic Power Station (NAPS) • To verify predictions of the code• To experimentally establish the onset of natural circulation in Indian

PHWRs

Tests conducted at

power levels of 0.956, 1.991 and 3.249 percent of full power (FP)

Initial Conditions

secondary side pressure of the SG : SG level : primary pressure :

14.6 bar12.9 m 86 bar

NATURAL CIRCULATION TESTS IN INDIAN PHWR


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Initial Conditions

For 0.956% FP and 1.991% FP tests :

1-1 mode of pump operation

0.956% FP tests :

For the 1.991% FP tests : Heat removal was by all the four SGs

Tests at 3.249% FP :

Heat removal was by two SGs

Initial PHTS flow rate, pressure and temperature corresponded to those of steady state thermosyphon at 1.991% FP

Tests at 0.956% and 1.991% FP : by switching off the PHTS pumps

Test at 3.249% FP :

increasing the power from 1.991% FP to 3.249% FP, when reactor was thermosyphoning at 1.991% FP

Test Procedures



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Test Procedures

Parameters recorded during the tests

• Channel flows and differential temperatures for 16 channels

• SG differential temperatures, pressure and level for all the four SGs

• Differential pressure across each RIH and the corresponding ROH

• PHTS pressure• Neutron power and• Channel outlet temperature for all the channels

PHTS pressure was maintained with the help of Fuelling Machine Supply Pump



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-30

-8

14

36

58

80

T-°

C

160.00 320.00 480.00 640.00 800.00

NORMAL FLOW DIRECTION : SOUTH - NORTH

Time (s)

N-13

Q-16

S-8

Typical transient behaviour, representative of most of the channels is shown by channel S-8 (Fig.3).

In channel N-13 delta-T indication was about –16.1ºC. i.e., 16.1ºC delta-T with flow reversal.

In channel Q-16 flow reversal has taken place twice.

In majority of the channels coolant was flowing in the forward direction with a few flowing in the reverse direction under steady state conditions.

None of the monitored channel delta-Ts showed negative values for the 1.991% and 3.249% FP tests indicating the absence of flow reversal.

Channel ΔT variation at 0.956%FP

RESULTS AND DISCUSSIONS


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A B C D E F G H J K L M N O P Q R S T A B C D E F G H J K L M N O P Q R S T A B C D E F G H J K L M N O P Q R S T2 0 0

2 1 0

Cha

nnel

Tem

pera

ture

( °

C)

T O P T O PB O T T O M T O PB O T T O M B O T T O M

1 2 1 9 2 0

AB

ST

R O W R O W R O W

2 2 0

2 3 0

2 4 02 4 0

2 5 0

2 6 0

2 7 0

2 8 0

1 0 C O L U M Nth

1 1 C O L U M Nth 1 2 C O L U M N

th

N -S F L O W D IR E C T IO NS -N F L O W D IR E C T IO N

3 .2 4 9 %F P

1 .9 9 1 %F P

Channel outlet temperature variation across different vertical columns for tests at 1.991 and 3.249% FP

Channels at the bottom of the calandria have a larger elevation difference.

Hence, flows in the lower channels are expected to be more.

This leads to lower outlet temperatures in lower channels.

This is all the more significant since the bottom channels generate more power than the corresponding top channels

RESULTS AND DISCUSSIONS


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Analysis was carried out using the computer code TINFLO-III Developed in-house

-solves the one-dimensional momentum and energy equations using the finite difference method

-Flow coastdown is modeled based on the test data

-Steam Generator secondary side is modeled by a single volume

-The code lumps all parallel flow paths in the half-core into one equivalent flow path.

- code can predict only the average core behaviour

METHOD OF ANALYSIS


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0.01

0.00

0.10

1.00

0.100.01 1.00 10.00 100 1000.00

TEST

PREDICTION

ESTIMATED STEADY STATE FLOWBASED ON HEAT BALANCE

Flow

- fra

ctio

n

T ime (s)

COMPARISON OF EXPERIMENTAL & THEORETICAL RESULTS

• The normalization of flow is with respect to 2-2 operation mode.

• A very good match during the initial coastdown phase.

• Beyond this, though showing similar trends, differ widely.

• This difference is attributed to the measurement inaccuracy at these low flows (less than10% of the full flow).

Variation of flow with time for test at 0.956% FP


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0.01

0.00

0.10

1.00

0.100.01 1.00 10.00 100 1000.00

TEST

PREDICTION

ESTIMATED STEADY STATE FLOWBASED ON HEAT BALANCE

Flow

- fra

ctio

n

T ime (s)Variation of flow rate for test at 1.991%FP



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Pres

sure

dro

p (k

Pa)

2 .0

2 0 0 .00

-1.0

5 .0

8 .0

1 1.0

1 4.0

4 0 0 .00 6 0 0 .00 8 0 0 .00 1 0 0 0.0

T im e (s)

P R E D IC T IO NT E S T

Variation of header to header delta-P for test at 1.991%FP

• Same trend • Time lag between the predicted and measured minimum and peak

pressure drops • At time t=o the delta–P indications are off-scale since these delta–

P indicators have a range of only –50 cm to + 250 cm and the delta-P is much higher than this (13000 cm) when the primary pumps are operating.

HEADER TO HEADER PRESSURE DROP


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0 .0

8 .0

Aver

age

Chan

nel

T (°

C)

2 80 .00

T im e (s)

16 .0

24 .0

32 .0

48 .0

5 60 .00 8 40 .00 1 120 .00 1 400 .00

P R E D IC T IO N

T E ST

0 .00

Variation of average channel delta-T for test at 0.956%FP

• The experimental curve is the mean of the seven channel delta-Ts • measured and predicted delta-Ts show the same trend • the predicted peak value is much larger (about 30%) than that

measured• There is a time lag between the peak values• Prediction is based on an equivalent feeder of average length



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16.0

200.00

T im e (s)

0.00

0.00

32.0

48.0

64.0

80.0

400.00 600.00 800.00 1000

PR ED ICTIONT EST

Aver

age

Chan

nel

T (°

C)

Variation of average channel delta-T for test at 1.991%FP

CHANNEL TEMPERATURE DIFFERENTIAL


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• The tests conducted in unit-1 of NAPS have confirmed the efficacy of natural circulation core cooling at decay power levels.

• Flow reversal occurred in some channels during the test at 0.956% FP.

• The channel elevation plays a significant role in determining the flow rate through the different channels.

• An interesting aspect of NAPS-1 test is the fact that the same unit of the same power station could be saved because of natural circulation core cooling in the event of an accident.

- A few years after the experiments, the unit faced a prolonged station blackout. ( even class-1 power supply was not available)

- Fire fighting water was injected by diesel pumps into the SG secondary side.

- Steam pressure was maintained in the secondary side using the Atmospheric SDVs.

- The main function of removal and transportation of heat from core to SG was achieved by natural circulation very efficiently and the unit could be saved.

CONCLUDING REMARKS

Documents

BARC IAEA Training Course/Workshop on Natural Circulation in Water Cooled Nuclear Power Plants, ICTP, Trieste, June 25-29,2007 Examples of Natural Circulation