the progress of i&c system development in small modular ...design...

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The Progress of I&C System

Development in Small Modular

Reactor ACP100

Chen Zhi

Nuclear Power Institute of China

22nd, May, 2013，Vienna

1. Background

2. The Main Technical Characteristics of ACP100 and Its

R&D Progress

3. The Basic Design Principles and the Overall Architecture

of ACP100 I&C System

4. The Preliminary Schemes of the Main I&C Systems of

the ACP100 Nuclear Steam Supply System

5. Summary

CONTENTS

1.1 The SMRs in the world

The sole development of large NPPs is not meet with the

extensive application demands of the huge electricity

generation and non-electricity applications.

More and more countries are developing the advanced small

and medium sized reactors (SMRs) to meet the more extensive

requirements.

13 SMRs are under construction in six countries and the

approximately 45 innovative SMR concepts research for

electricity generation and other applications is being carried

out.

Background

1.2 The requirements for the SMRs in China

The requirement in electricity

generation area.

SMR will be the best choice for vast

inland areas and outlying areas of China.

Background

1.2 The requirements for the SMRs in China (Continued)

The requirement in industrial and

process heat supply area.

The approximately 900 millions

tons of industrial steams are

consumed in China every year. The

emission of greenhouse gases in

the course of producing these

steams occupies 10% of the total

emission in China.

Background


The requirement in desalination

area.

The fresh water sources are very lack

in China. Most of industries locate in

coastal areas, the serious lack of

fresh water resources having become

the bottleneck.

Background


The requirement in city heat supply

area.

The energy demands of city heat

become larger and lager in north cities

of China, especially in “Three North”

regions. The pollution condition is

worse than before with the large

consumption of fossil fuel due to the

increasing demands of heat supply.

Background

1.3 What is the ACP100?

Considering the above background, China

National Nuclear Corporation (CNNC) is

carrying out the development of the small

modular reactor, which is coded ACP100 .

It is an innovative PWR based on existing

PWR technology , adapting “passive” safety

system and “integrated” reactor

design technology.

Background

ACP100

1. Background

2. Main Technical Characteristics and R&D Progress

of ACP100

3. Basic Design Principles and the Overall

Architecture of ACP100 I&C System

4. The Preliminary Schemes of the Main I&C

Systems of the ACP100 NSSS

5. Summary

CONTENTS

2.1 Main technical parameters of ACP100

Main Technical Characteristics

Type Integrated PWR

Thermal power 310 MWt

Electrical power ~100 MWe

Design life 60 years

Refueling period 2 years

Coolant average temperature 303 ℃

Operation pressure 15MPa（a）

Fuel active section height 2150 ㎜

2.1 Main design parameters of ACP100 (Continued)


Fuel assembly number 57

Drive mechanism type magnetism lifting

Control rod number 25

Reactivity control method Control rod、solid burnable

poison and boron

Steam generator type OTSG

Main steam pressure 4MPa（a）

Main pump type canned pump SSE level ground seismic peak acceleration

0.3g

2.2 Main technical characteristics of ACP100


Integrated layout of primary

system and equipment.

So the large LOCA accident is eliminated.

And the dimension and the amount of the

penetration in RPV can be also reduced.

Large primary coolant inventory.

The thermal inertia is increased.

Small radioactivity storage

quantity. ACP100

2.2 Main technical characteristics of ACP100(Continued)


The layout of RPV and

equipment is benefit for natural

circulation.

Smaller decay thermal power. And it is easier to achieve safety by

the way of “passive”.

Reactor and spent fuel pool

are laid under the ground level.

So it is better to withstand exterior

accident and good for the reduction

of radioactive material release.

2.3 Research and development progress


Design work.

Standard design, is completed by the end of 2012.

The preliminary safety analysis report (PSAR) is also

finished.

2.3 Research and development progress


Testing and verification aspects.

The six test research subjects, including control rod

drive line anti-earthquake test, passive emergency core

cooling system integration test, etc., is planed to

completed in 2013.

Thermal hydraulic testing hall Passive emergency core cooling system

2.3 Research and development progress (Continued)


Licensing.

The contract of ACP100 combined research with

National Nuclear & Radiation Safety Center (NNRSC)

was signed in 2011. And several specific research

programmers and standard design safety analysis

combined research with NNRSC will be carried out in

year 2013.

2.3 Research and development progress (Continued)


Site selection.

The demonstration ACP100 nuclear power plant, with

two 310Mwth reactors, will be located in Putian City,

Fujian Province in the east coast area of China.

ACP100

Demonstration Site

Expecting construction

in June, 2014

1. Background

2. Main Technical Characteristics and R&D

Progress



4. Preliminary Schemes of the Main I&C

systems of ACP100 NSSS

5. Summary

CONTENTS

Meet the defense-in-depth concept

Compliance with the single failure criterion

Diversity design of I&C system

Basic Design Principles

Meet the defense-in-depth concept

During the normal operation and operational transients, the plant

control systems will act to maintain and restore the plant normal

operation;

The reactor protection system will act to limit the consequences

of any anticipated transient of malfunction;

The reactor protection system will initiate selected protective

functions to mitigate the consequences of design basis events;

During the serious accidents, providing the serious accident

monitoring and control function to limit the consequences of core

melting and radioactive substances releasing.


Compliance with the single failure criterion

RPS

individual inputs and the logic outputs of the reactor trip: 4 divisions.

Engineered Safety Feature Actuation System (ESFAS): also consists of

individual divisions;

The supporting systems (Power supply)

follow the same redundancy design criteria

the 4 protection division sets are power supplied by 4 independent

the diverse actuation system is supplied by 2 independent UPS.


Diversity design of I&C system

The different hardware and software platform will be applied to

the safety and non-safety I&C systems separately.

The digital RPS is designed according to functional diversity.

For the CMF of the RPS, the diverse actuation system provides

corresponding protection.


I&C system overall architecture

Overall Architecture

process interface level

Control and

protection level

Operation and

information level

Site management level

1. Background


Progress



4. Preliminary Schemes of the Main I&C systems of

ACP100 NSSS

5. Summary

CONTENTS

4.1 System configuraion

System Configuration

I&C systems of ACP100 Nuclear Steam Supply

System (NSSS) include:

Reactor nuclear instrumentation system

Reactor protection system

Diverse actuation system

Reactor control system

Rod control and rod position monitoring system

Reactor in-core instrumentation system

Loose parts and vibration monitoring system

Other process control systems

4.2 The general control scheme

General Control Scheme

Automatic control combined with the manual control

Reactor power control system and the steam generator (SG)

feedwater control system use the manual scheme instead of

automatic scheme below 20% full power.

The other control systems have an automatic control range

from 0 to 100% full power.

The systems are designed to return automatically to

equilibrium conditions following the load variations in steps of

±10% FP transient and load variations in continuous ramps

with a gradient of ±5%FP per minute transient.

4.3.1 reactor nuclear instrumentation system

System Preliminary scheme

system function

continuously monitor reactor power

signals are used for control and protection

block automatic and manual rod withdrawal when neutron

flux exceeds the interlock setpoint

initiate reactor trip on high nuclear flux and high neutron flux

variation


Systems Preliminary scheme

system scheme

(A) source range

consists of 4 identical and independent channels which

furnish redundant neutron flux signals during shutdown and

initial plant startup.

Detector covers a flux range from 11-9 %FP to 10-3 %FP.

(B) intermediate range

consists of 4 identical and independent channels which

furnish redundant neutron flux signals.

Detector covers a range from about 10-6％FP to 100％FP.



system scheme

(C) power range

consists of 4 identical and

independent channels which

furnish redundant neutron flux

signals.

Detector covers a flux range

from 10-6％FP to 200％FP.

The all detectors are arranged in

the 12 independent holes outside

the RPV. The cabinets include four

safety protection cabinets and a

non-safety control cabinet.

0¡ã

22.5¡ã

45¡ã

90¡ã112.5¡ã

135¡ã

180¡ã

202.5¡ã

225¡ã

270¡ã 292.5¡ã

315¡ã

:sourge range(proportional counter)

:intermediate range(compensated ion chamber)

:power range(uncompensated ion chamber)

4.3.2 SG feedwater control system


system function

maintain the OTSG secondary side pressure at a constant

make the feedwater flow accommodate the load requirements



system scheme

(A) During 0 to 20%Pn, the feedwater flow is

controlled by manual

The bypass control valves are used to adjust the feedwater

flow to the steam generator manually through the startup

feedwater line and the startup feedwater pump.

(B) During the 20%Pn to 100%Pn

The feedwater flow is regulated automatically through the

main feedwater pump, main feedwater control valve and the

main feedwater line. The bypass feedwater channel is not

operation at this condition.



system scheme （Continued)

The automatic control of the feedwater flow is accomplished

by the main feedwater flow control valve in conjunction with the

feedwater pump speed control.

pump speed controller

¡ ÷P across the main valve

¡ ÷P setpoint

feedwater valve controller

SG steam pressure

feedwater flow

pump speed

feedwater valve position

steam flow

4.3.3 reactor power control system


system function

maintain the reactor coolant average temperature to a

constant under steady state conditions .

enable the nuclear plant to accept a step load increase or

decrease of 10% FP and a ramp increase or decrease of 5%FP

per minute within the given load range without reactor trip or

the steam dump system actuation.

4.3.3 reactor power control system


system scheme

(A) the temperature channel

compares the coolant average temperature measurement with

the reference temperature (a fixed value). The error between

them is the primary control signal of the RPC system.

(B) the power mismatch channel

receives the nuclear power signal and the total feed water

flow of the secondary side.

The error of the power mismatch channel is added to the temperature

error signal. The final error signal is processed by the rod speed program

which produces the control rod travel speed signal and two direction logic

signals.

4.3.4 reactor in-core instrumentation system


system function

The reactor in-core instrumentation (RII) system is made up of

the neutron flux measurement subsystem

the in-core temperature measurement subsystem

the vessel level measurement subsystem.



system function

1) The neutron flux measurement subsystem:

measuring the real time core neutron flux and making flux

map;

providing the necessary inputs to the core online monitoring

system and other data processing system;

combining the other condition data of the reactor to check

the calibration of the ex-core nuclear instrumentation.



system function

2) The in-core temperature measurement subsystem :

provides the temperature values of the reactor cooling water at

the fuel assembly outlet.

to calculate the reactor coolant maximum temperature, average

temperature and the minimum core saturation margin.

3) The vessel level measurement subsystem

provides the information whether the key points in the vessel is

submerged or not.



system scheme

The all detectors are inserted into the core through the upper

section of the RPV.

The in-core neutron detector is self-power detector. The SPND and

the thermocouple are integrated to the neutron and temperature

detectors assemble.

The in-core temperature measurement subsystem complies with

the redundant principle. All equipments, including the Inadequate

Core Cooling Monitoring (ICCM) cabinets, are divided into train A

and train B, and are physically and electrically separated.

The vessel level measurement subsystem also complies with the

redundant principle and is divided into train A and train B.



system scheme

The layout of the detectors of the RII system.

Thermocouples(TRAIN A)

Thermocouples(TRAIN B)

Level detectors(TRAIN A)

Level detectors(TRAIN B)

SPND layout Thermocouple and level

detectors layout

4.3.5 reactor protection system


system function

monitor the physical parameters which are essential to the

reactor safety

detect any transient changes in these parameters and

triggered as required the operation of safety systems.

limit the consequences of any accident conditions and then

to ensure safety of the reactor.



system scheme

The ACP100 RPS has a configuration of four redundant divisions.

Reactor trip

Four redundant measurements, using four separate sensors.

A partial trip signal for a parameter is generated if one channel’s

measurement exceeds its predetermined or calculated limit.

Each division sends its partial trip status to each of the other three

divisions over isolated data links.

Each division is capable of generating a reactor trip signal if two or

more of the redundant channels of a single variable are in the partial trip

state.

There are eight reactor trip switchgear breakers. Each of the four

reactor trip divisions consists of two reactor trip circuit breakers. The

reactor is tripped when two or more actuation divisions output a reactor

trip signal. The reactor trip operates upon loss of voltage.



system scheme

Engineered Safety Feature Actuation System Four sensors normally monitor each variable used for an ESF

actuation.

When the measurement exceeds the setpoint, the output of

the comparison results in a channel partial trip condition.

The signals are combined within each division of ESF

coincidence logic to generate a system-level signal.

The ESF actuation operates on presence of the voltage.



ACP100 RPS architecture

4.3.6 Diverse actuation system


system function

The diverse actuation system (DAS) is a non-safety system

that provides a diverse backup to the reactor protection.

DAS monitors some plant variables, accomplishing reactor

trip and ESF actuation in the case of the CMF occurrence in the

RPS system.



system scheme

The diverse actuation system uses the different I&C platform

from that being used by the reactor protection system.

The diverse actuation system uses sensors that are separate

from those being used by the protection system.

Actuation interfaces are different between the diverse

actuation system and the protection system.



system scheme (continued)

The signal from the sensor is

hardwired to the DAS system.

The measurements are compared

against the setpoints for the diverse

actuation to be generated.

When the measurement exceeds the

setpoint, the output of the comparison

results in a channel partial trip condition.

As a diverse design, the reactor trip

signal generated by the DAS will be

transmitted to the CRDM power supply

cabinet instead of the reactor trip

breaker.

DAS architecture

1. Background


Progress


Architecture

4. Preliminary Schemes of the Main I&C systems

5. Summary

CONTENTS

ACP100 uses the “passive” safety technology and

“integrated” reactor technology and has the high

intrinsic safety.

According to the primary system design, ACP100

I&C system standard design is completed by the

institutes of CNNC.

The general I&C architecture, main control room

design, electricity supply plan, I&C equipment

design, etc., will be optimized in the next design

works aiming at the demonstration projects.

Summary

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