Technological challenges and development of ...Design Specific/AHWR/Presentations/2013... · instrumentation and electrical systems located inside the containment: ... No of Down

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Technological challenges and development of instrumentation sensors and techniques

for Indian Advanced Heavy Water Reactor (AHWR)

Rajalakshmi.R Bhabha Atomic Research Centre

Mumbai, INDIA

Technical Meeting on “Instrumentation and Control in Advanced Small and Medium-sized Reactors (SMRs)”

21 - 24 May 2013 VIC, Vienna, Austria

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The Indian Advanced Heavy Water Reactor (AHWR)

Indian AHWR is a 300 MWe, vertical, pressure tube type, heavy water moderated and boiling light water cooled natural circulation reactor using (Th-233U )MOX and (Th-Pu) MOX fuel. (An innovative configuration that can provide low risk nuclear energy using available technologies)

Safety and security – Indian Technological approach

No unacceptable radiological impact outside the plant boundary with (a)Failure of all active systems, and

(b)Failure of external infrastructure to provide coolant, power and other services, and

(c)Malevolent acts by an insider, one of the consequences of which is the failure of instrumentation signal initiated shutdown actions, and

(d)Inability of plant operators to manage the events and their consequences, for a significantly long time.

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The Indian Advanced Heavy Water Reactor (AHWR)

Safety and security – Indian Technological approach

A robust structural containment to protect internal systems against external threats.

Even with (a) non-availability of all services and functions located outside the containment, including control room functions and operator actions, and

(b) non-functionality or compromised functionality of any instrumentation and electrical systems located inside the containment:

Capability for safe shutdown of the reactor

Availability of an adequate capacity heat sink, either inside the containment, or accessible across containment structure.

Natural circulation driven transfer of decay heat to the heat sink.

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Advanced Heavy Water Reactor (AHWR) - Salient features

Major design objectives

Significant portion of energy from thorium. 65% of power from Th

Fuel (Th-U233)O2 & (Th-Pu)O2

Reducing probability of severe accidents

Several passive features

10 days grace period

No radiological impact in public domain

Passive shutdown system to address various threat scenarios.

Design life of 100 years.

Easily replaceable coolant channels.

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Advanced Heavy Water Reactor (AHWR) Innovative passive technologies

Core heat removal by natural circulation of coolant during normal operation and shut down condition.

Slightly negative void coefficient of reactivity.

Passive safety systems working on natural laws.

Large heat sink in the form of Gravity Driven Water Pool with an inventory of 8000 m3 of water, located near the top of Reactor Building.

Injection of cooling water by Emergency Core Cooling System directly inside the fuel cluster.

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Advanced Heavy Water Reactor (AHWR) Innovative passive technologies

Containment cooling by passive containment coolers.

Passive containment isolation by water seal during LOCA.

Two independent shutdown systems (primary and secondary).

Passive poison injection in moderator in the event of non-availability of both primary as well as secondary shut down system due to failure or malevolent insider action.

Facilitate siting close to population centers.

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Defence in Depth (IAEA-TECDOC-1434) → 1. Prevention of abnormal operation and failures

Enhance prevention by increased emphasis on inherently safe design characteristics and passive safety features, and by further reducing human actions in the routine operation of the plant.

→ 2. Control of abnormal operation and detection of failures

Give priority to advanced control and monitoring systems with enhanced reliability, intelligence and the ability to anticipate and compensate abnormal transients.

→ 3. Control of accidents within the design basis

Optimized combination of active & passive design features; limit consequences such as fuel failures; large heat sinks within/ around containment, minimize reliance on human intervention by increasing grace period, e.g. between several hours and several days.

Safety Criteria for Innovative Advanced Reactor Designs

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Defence in Depth (IAEA-TECDOC-1434)

→ 4. Control of severe plant conditions, including prevention and mitigation of the consequences of severe accidents.

Increase reliability and capability of systems to control and monitor complex accident sequences; decrease expected frequency of severe plant conditions; e.g. for reactors, reduce severe core damage frequency by at least one order of magnitude relative to existing plants and designs, and even more for urban-sited facilities.

→ 5. Mitigation of radiological consequences of significant releases of radioactive materials

Avoid the necessity for evacuation or relocation measures outside the plant site.

More independence of levels from each other.

Safety Criteria for Innovative Advanced Reactor Designs

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Defence-In-Depth (DID) of AHWR

• Level 1 DID: – Elimination of the hazard of loss of coolant flow :

• Heat removal from the core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant.

– Reduction of the extent of overpower transient : • Slightly negative void co-efficient of reactivity. • Low core power density. • Negative fuel temperature coefficient of reactivity. • Low excess reactivity

• Level 2 DID: – Control of abnormal operation and detection of failure

• An increased reliability of the control system achieved with the use of high reliability digital control using advanced information technology.

• Increased operator reliability achieved with the use of advanced displays and diagnostics using artificial intelligence and expert systems.

• Large coolant inventory in the main coolant system.

10

Passive Systems in Defence-In-Depth of AHWR

• Level 3 DID:

– Control of accidents within the design basis

• Increased reliability of the ECC system, achieved through passive injection of cooling water directly into a fuel cluster through four independent parallel trains.

• Increased reliability of a shutdown, achieved by providing two independent shutdown systems. Further enhanced reliability of the shutdown, achieved by providing a passive shutdown device

• Increased reliability of decay heat removal, achieved through a passive decay heat removal system, which transfers the decay heat to GDWP by natural circulation.

• Large inventory of water inside the containment (about 8000 m3 of water in the GDWP) provides a prolonged core cooling meeting the requirement of grace period.

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Passive Systems in Defence-In-Depth of AHWR

• Level 4 DID: – Control of severe plant conditions, including prevention of accident

progression and mitigation of consequences of severe accidents

• Use of moderator as heat sink.

• Flooding of reactor cavity following a LOCA.

• Level 5 DID: – Mitigation of radiological consequences of significant release of radioactive

materials

• The following features help in passively bringing down the containment pressure and eliminates any releases from the containment :

– Double containment;

– Passive containment isolation;

– Vapour suppression in GDWP;

– Passive containment cooling.

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Heat removal from core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant.

Some important passive safety features of

AHWR –1/4

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Passive injection of cooling water, initially from accumulator and later from the overhead GDWP, directly into fuel cluster.

(Th-Pu) MOX

Fuel pins

(Th-233U) MOX

Fuel pins Central Tube for

ECCS water

AHWR FUEL CLUSTER

Passive Containment isolation

Passive Containment Cooling

Some important passive safety features of

AHWR –2/4

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14

Some important passive safety features of AHWR –3/4

Passive Poison Injection System actuates during very low probability event of failure of wired shutdown systems (SDS#1 & SDS#2) and non-availability of Main condenser

Passive Poison Injection in moderator during overpressure transient


Use of moderator as heat sink Water in

calandria vault

Flooding of reactor cavity following LOCA


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AHWR- Main Heat Transport System (MHT)

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Design Parameters

Common Inlet header 1

No of Feeders/ Channels 452

No of Steam Drum 4

No of Tail pipes connected to each steam drum. 113

No of Down comers from each steam drum connected to common inlet header 4

Fuel Burn Up 34,000 Mwd/Te

Core Flow rate 2,306 Kg/Sec

Coolant inlet temp 261.4 ºC

Feed water temp 130 ºC

Average Quality 17.6%

Steam Pressure & Temp 70 bar & 285 ºC

AHWR- I&C Architecture

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Regulation of reactor power and process variables of the plant within limits

Control and monitoring of

* Main Heat Transport System and auxiliary process systems

* Turbine and turbine auxiliary systems

* Generator and its auxiliary systems

* Electrical systems

* Station common process and service systems

Ensuring safety of the plant under all operating conditions and during all postulated emergency situations and

Sensing accident conditions and initiating the operation of systems required to mitigate the consequences of such accidents.

I & C systems of AHWR are designed to carry out the following important parameters

Based on their functions C&I systems are classified as safety systems (IA), safety related systems (IB)and systems not important to nuclear safety (NINS).


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Safe Shut down of reactor

Decay heat removal

Protection of coolant boundary against over pressure

Prevent / minimise release of radio activity

Control & monitoring of plant parameters

Maintain plant in safe state

I & C systems of AHWR perform following important safety functions

Design principles followed:

Redundancy

Diversity

Physical separation

Fail Safe & Single failure criterion

Fault Tolerance

Online Testability & Maintainability


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Group-1 - SDS-1 - ECCS - CLMS-BSF - Plant systems for normal operation & safety related control systems

Two- Group Philosophy:

To enhance the Reliability, I&C systems are divided into two groups namely Group-1 & Group-2

Each group has the capability to perform the safety functions independent of the systems in other group

I&C hardware are independent with respect to sensors, power supplies, cables and are physically separated

Functional independence of the two groups ensure single local event (Fire or Pipe failure etc) will not result in multiple component or system failure

Group-2

- SDS-II - CIS

AHWR- Technological Challenges

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Block Diagram Of C&I Architecture of AHWR


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Triple Modular Redundent Architecture


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Dual System with Hot Standby

AHWR- Technological Challenges

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The technological challenges in AHWR are

To design and develop novel devices suitable for precision measurement of bi-directional low velocity flow in feeders and identify the flow reversal in the channels

To develop sensors for void measurements and an algorithm to compute channel quality and hence the channel power from this void fraction

To identify the occurrence of stagnation channel break

Trip the reactor if any of these parameters exceeds the safe limit.

The main design requirement of AHWR is to maintain adequate thermal and stability margin. Instability and Critical Heat Flux (CHF) in the system should be avoided in the operation under all operational states and transients. Hence the channel flow and channel power monitoring instrumentation becomes very important for channel safety and integrity.

AHWR-R&D and Experimentation

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Development of Instrumentation sensors and methods and their qualification in experimental facilities simulating AHWR operating conditions were taken up for

The measurement of low velocity Bi-directional flow

void fraction

Two-phase flow

Measurement technique for channel power

Development of Bi-directional venturi for AHWR

Channel Flow Measurement

Channel flow monitoring instrumentation is very important in

AHWR for channel safety & Integrity

Requirements: To measure low flow velocities

Minimum permanent pressure loss

To identify flow reversal

Flow measurement in high temperature and high pressure natural circulation system

Channel flow measurement in all 452 inlet feeder lines

to monitor

1) Fwd flow during reactor normal operation

2) Rev flow during (for eg: refuelling/ defuelling, Feeder LOCA)

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The objective is to provide the novel device suitable for precision measurement of bi-directional low velocity flow and identify the flow reversal when fluid is at high temperature and high pressure AHWR operating condition.

Design Basis:

Why Venturi flow element?

Minimum permanent pressure loss (since pressure loss affects Natural Circulation flow).

Minimum straight lengths requirement.

Bi-directional flow measurement is feasible by selecting same convergence and divergence angles.

Development of Bi-directional venturi for AHWR

Channel Flow Measurement

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Design Of Bidirectional Venturi

DP1 DP2

P1 P2 P3

DP1 = P1 - P2 DP2 = P3 - P2

D = Pipe diameter, m

d = throat diameter, m

= beta ratio of Bidirectional Venturi

= Venturi discharge coefficient

= Differential pressure pa

= Density of water in

= Flow rate in

dC

p

Q 3mkg

sm3

Schematic of Bidirectional venturi, beta = 0.5

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BIDIRECTIONAL VENTURI FLOWMETER

Low permanent head loss

Good measurement sensitivity at low flow velocities

Reasonably constant discharge co-

efficient for wide range of Reynolds numbers

Flow reversal identification

Bidirectional Venturi

Salient features: Bi-directional flow measurement

Accuracy better than ± 0.5% (Calibrated at room temperatures and the characteristics are stored for better accuracy)

Sensor shall introduce least disturbance in the process

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For orifice, venturi and nozzle design, empirical standards such as BS1042 is used.

For Bidirectional venturi design, no such standard is available.

Hence systematic design approach for dimensional optimization by theoretical modeling and simulation studies using Computational fluid dynamic (CFD) code were taken up.

Experimental & CFD Results were compared


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Pressure and velocity profile

Working

fluid

Temperature

(0C)

Pressure

(bar)

Density

(kg/m3)

Viscosity

(Pa s)

Water 30 1 1,000 1.030e-3

Velocity profile for BDV

beta =0.5, cone angle:30 deg

Pressure profile for BDV

beta =0.5, cone angle:30 deg


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CFD simulation results

0 100 200 300 400 500 600-100

0

100

200

300

400

500

600

700

800

DIF

FE

RE

NT

IAL

PR

ES

SU

RE

(m

mW

C)

FLOW RATE (LPM)

DP1, Cone angle = 21 deg.




0 100 200 300 400 500 600

0.10

0.15

0.20

0.25

0.30

0.35

PR

ES

SU

RE

LO

SS

(% O

f DP

)

FLOW RATE

Cone angle = 21 degree

Cone angle = 30 degree

0.0 2.0x104

4.0x104

6.0x104

8.0x104

1.0x105

0.4

0.5

0.6

0.7

0.8

0.9

1.0

DIS

CH

AR

GE

CO

EF

FIC

IEN

T

REYNOLDS NUMBER

Cone angle = 21 deg.

Cone angle = 30 deg.


(2) Pressure loss characteristics

(1) Flow Characteristics

(3) Discharge co-efficient characteristic

100 200 300 400 500 600

0

100

200

300

400

500

600

700

Dif

fere

nti

al P

ress

ure

(m

mW

C)

FLOW RATE (lpm)



Experimental characteristics

(4) Flow Characteristics

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Bidirectional Venturi Experimental Results

0 100 200 300 400 500 600-100

0

100

200

300

400

500

600

700

800

Dif

fere

nti

al P

ressu

re (

mm

WC

)

FLOW RATE (lpm)

Expt. pressure drop, Cone angle = 30 deg.

CFD pressure drop, Cone angle = 30 deg.

0 1x104

2x104

3x104

4x104

5x104

6x104

7x104

8x104

9x104

1x105

1x105

1x105

0.5

0.6

0.7

0.8

0.9

1.0

DIS

CH

AR

GE

CO

EF

FIC

IEN

T

REYNOLDS NUMBER

Experimental Cd, Cone angle = 30 deg.

CFD studies Cd, Cone angle = 30 deg.

Comparison: CFD & Experimental Pressure drop characteristics

Comparison: CFD & Experimental Discharge coefficient characteristics

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Uncertainty in )1

(

n

nsqrtUcd

Bidirectional Venturi Uncertainty Analysis

Maximum deviation of CFD & Experimental results is 3.7%.

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Development of Two-phase Flow Instrumentation

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Measurement of two-phase flow in steam-water natural circulation system is complex and various attempts were made to understand and develop instrumentation so that channel power can be obtained for effective channel monitoring.

OBJECTIVE:

Design and Development of Two-Phase Flow Sensor for High Pressure Steam-Water Applications

1. Impedance based void sensor (Rotating Electric field Admittance probe)

2. Two-phase mass flux measurements by pitot tubes

assembly and Gamma Ray Densitometer

Rotating Electric Field Admittance Probe

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Methodology :

Mixture impedance relative to that of the separate phases gives

void fraction.

Transit time of random flow fluctuations to travel between

adjacent sensors can determine the fluid phase velocities.

Basic principle of measurement: The basic principle of the sensor developed is based on the difference in electrical properties between the liquid and vapor phases, the total admittance i.e. capacitance and resistance relative to that of the separate phases gives void fraction.

)( j

mixmixmixeYj CGY


5/22/2013 36

Sensor shall not introduce disturbance to the flow and it shall not cause pressure drop in the system as these parameters affects the driving force available in natural circulation systems

Compensation for conductivity variation due to fluid temperature and ionic concentration changes

Prediction of various flow regimes in two-phase mixture through the measurement data

End Connection

S.S Pipe

M.I Cable

Ceramic

Insulation

Electrode Plate

Sensor Construction Details

Design Criteria


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The electrodes are mounted coaxially inside to form part of pipe wall to eliminate the disturbance to the flow.

The sensor causes no pressure drop in the system.

The electric field which is perpendicular to the flow is rotated electronically to distribute it through out the sensor volume.

The conductivity changes due to temperature and ion concentration are compensated by measuring relative admittance w.r.t. reference sensor located in single phase region.

The phase shift of sensor signal with respect to excitation voltage determines flow regimes.

Salient features :

5/22/2013 38

Steam Water Experimental Results

Development of prototype

sensor for 100 bar pressure

and 220 °C temp for steam-

water ambience was

completed and

commissioned successfully

in Parallel Channel natural

circulation experimental

facility.

Qualification Experiments

were carried out at various

loop pressures and heater

powers.


Uncertainty Analysis

• Uncertainty was estimated for a set of Air-Water experimental data and analysed for the various flow regimes using data obtained from the air-water visual experiments by signature analysis.

• Uncertainty was found to be within 3 to 9% for the various flow regimes from bubbly to churn flow upto 80% void fraction.

Advantages :

• Simple data interpretation for real time measurements.

• Fast response, hence can be used for transient studies.

• Volume average technique for better results.

• Provides flow pattern discrimination.

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Two Phase Mass Flux Measurement by Pitot Tubes assembly and Traversing Gamma Ray Densitometer

Methodology

Cross section averaged Two Phase mass velocity measurements using

• Pitot Tubes assembly

Chordal void fraction and average mixture density measurements using

• Traversing Gamma Ray Densitometer

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High Pressure Natural Circulation Loop (HPNCL)

Heater

Gamma Source

Gamma Detector

Water

Steam + water

Condenser Steam Drum


Pitot tubes assembly

Gammm ray densitometer

Pitot Tubes assembly & Gamma ray densitometer in Experimental set up

Pitot Tubes Assembly

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1

2

3

4

5

Steam + water

Pipe cross section

0 mm

30 mm

60 mm

Scintillation Detector

PM tube

Electronics & single channel analyzer

Traversing gamma source

Traversing gamma detector

1

2

3

4

5

Void fraction measurement by Traversing

Gamma Ray Densitometer

5/22/2013 42

Void Fraction & Mass Flux Profile across Pipe cross Section

0

0.2

0.4

0.6

0.8

1

0 10 20 30 40 50

Distance from bottom of pipe (mm)

Void

fra

ction / M

ass F

lux-G

(K

g/m

-sec) G 30KW

G 40 KW

G 50 KW

VF 30 KW

VF 40 KW

VF 50 KW

Pressure = 30 Bar

VELOCITY PROFILES ACROSS THE VERTICAL CHORD

0

1

2

3

4

5

6

0 10 20 30 40 50

Distance from the bottom of pipe (mm)V

elo

city (

m/s

ec)

Velocity 30 KW

Velocity 40 KW

Velocity 50 KW

Pressure = 30 Bar

Rugged, simple, reliable and easy to operate

The data interpretation is simple.

Sturdy sensor for adverse high temperature

and high pressure steam water applications.

Introduces only very little disturbance in low path, which is very important for two phase natural circulation studies.

Advantages


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44

Integral Test Loop (ITL) for Experimental Demonstration

44

ITL is a scaled facility which simulates the MHT, ECCS, IC System along with the associated controls of the AHWR

The integral or global scaling is based on the power-to-volume scaling philosophy.

All these sensors namely bi-directional venturi, rotating electric field admittance probe and multi beam gamma ray densitometer are planned to be installed in this facility for experimental qualification under reactor operating process conditions.

BDV Installed in ITL

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45

Technique for Channel Power Measurement

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Algorithm evolved for channel power computation using two-phase pressure drop measurements in the vertical tail pipes of AHWR.

The pressure drop (P) measured in the vertical tail pipes be considered to be comprising of gravitational pressure drop component only

The measured P used to obtain the density of the steam water mixture coming out of fuel channels and in turn void fraction in the pipe section. Steam quality then obtained in each channel using appropriate experimentally verified correlation.

Power in channel estimated using the measured channel mass flow rate and channel inlet/outlet enthalpy.

The uncertainty in measurements will be established through detailed experimentation.

46

Channel stagnation break identification

46 5/22/2013

In Natural circulation based Boiling Water Reactors small break LOCA detection is important.

Single channel event which is of concern for safety in these pressure tube type reactors is, stagnation channel break which is typically a small inlet feeder break of specific size.

The consequences such as fuel over heating in that channel can be avoided if there is a trip on channel low flow.

Since flow trip in all 452 channels is difficult, detection of steam leak is essential to detect this LOCA of various sizes.

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Channel stagnation break identification

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For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault.

To identify this channel stagnation break occurrence, acoustic sensors are provided in the reactor vault.

The outputs of these sensors are analyzed for noise level and spectrum characteristics to detect the steam leakage.

The detection technique is reliable and plays a very important role in ensuring safety of the reactor.

Reactor Trip

The signals from these acoustic sensors, flow measurements in inlet feeders and void fraction measurements in tail pipes generate reactor trip signal.

48 48 5/22/2013

Concluding Remarks

The technical challenges in I&C measurements for pressure tube type AHWR and general I&C architecture addressed.

The channel flow, channel power and channel stagnation break identification - important monitoring instrumentation for channel safety.

All the 452 channels in AHWR are instrumented with in-house developed bi-directional venturi flow meters for monitoring the forward flow and identify flow reversal if any during defueling/refueling.

The in-house developments of sensors for void fraction measurement are taken up and their qualification in high temperature, high pressure R&D experimental facility is in progress.

The channel power measurement technique was evolved and is monitored in all 452 channels.

A robust technique of steam leak detection is developed by acoustics method for channel stagnation break identification.

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Documents

Technological challenges and development of ...Design Specific/AHWR/Presentations/2013... · instrumentation and electrical systems located inside the containment: ... No of Down