china institute of atomic energy china national … institute of atomic energy china national...

43
China Institute of Atomic Energy China National Nuclear Corporation 2017-9-14 資料2

Upload: hangoc

Post on 31-Aug-2018

225 views

Category:

Documents


0 download

TRANSCRIPT

China Institute of Atomic Energy

China National Nuclear Corporation

2017-9-14

資料2

1. Overview about nuclear energy of China

2. SFR Development

3. Other Gen-IV Reactor Development

4. Conclusion

1

2

• China primary energy consumption is about 4.18 billion tons of standard coal in 2016, decrease0.79%。

• Crude oil, natural gas, and non-coal electricity consumption amount is about 109.8 million tons ofstandard coal. Coal consumption fell by about 143.6 million tons of standard coal.

Energy type Electricity

generation (*1012kW.h)

Ratio Increase

(compare to 2015)

Coal 4.3958 74.4% 2.6%

Hydro 1.0518 17.8% 5.9%

Wind 0.2113 3.5% 19.0%

Solar 0.0394 0.7% 33.8%

Nuclear 0.2127 3.6% 24.1%

Total 5.9111 4.5%

3

• The NPPs unites under operation is 35 (not include Taiwan region) till 31 December, 2016. And another 19 unites are under construction.

• The nuclear capacity is 33.6GW which is about 2.04% of total capacity.

4

• In 2016, China's electricity industry generated a total of 5.91 trillion kilowatt-hours, including 212.7 billion kilowatt-hours of nuclear power. And the nuclear grows 24.1% from 2015.

• The average utilization of nuclear power plant in China has declined for three years. In 2014, the average utilization rate of 22 nuclear power plants in China was 86.32%. In 2015, the NPP units increased to 28, and the average utilization rate dropped to 83.3%. In 2016, the NPP units reached to 35, but the average utilization rate fell to 79.55%. The operation time is 6,987 hours a year, which was nearly 300 hours shorter than the previous year.

• The reason are mainly: the national economy is in a period of adjustment; balance of the energy resource adjustment.

Electricity generation in 2016

5

• Energy policy – National Plan for Coping with Climate Change, to guarantee the

realization of the target of cutting the carbon emission intensity by 40 to 45 percent by 2020 from the 2005 level.

• Nuclear energy policy – The nuclear energy development should be sustainability, safe and with

high efficiency.

– Could provide a large scale nuclear power plant capacity in a limited time.

– The high level radioactive waste should be minimized.

– The roadmap of the nuclear development is tree steps: thermal reactor, fast reactor and fusion.

– The strategy of nuclear fuel cycle should be : the Closed fuel cycle based fast reactor.

6

• The Electric Power Development “13th five years” plan (2016-2020) was issued in 2016. The plan indicate that nuclear power will be put into operation with the capacity about 30GW and another 30GW will be under construction. The total capacity will reach to 58GW till 2020.

• The studies of the Chinese Academy of Sciences (CAS) show that the nuclear power installation will reach to 200GW by 2030, and more than 400GW by 2050.

45%

10% 10%

15%

19%

Energy structure in 2050

Coal

Oil

Gas

Nuclear

Renewable

7

8

Experimental Fast Reactor

CEFR Demonstration Fast Reactor

CFR600 Commercial Fast Reactor

CFR1200

2011 ~ 2023 ~2030

safety theory certification Scientific certification Fuel and material study Training and experience feed back

Industry scale to demonstration the closed fuel cycle

Safety Verification of large size SFR

Master of the large size SFR design and construction technology

Economic certification of large size SFR in China

Commercial operation Breeding nuclear fuel industrially Serially developing

9

Phases CEFR CDFR CCFR

Power(MWe) 20 600 ≥1000

Coolant Sodium Sodium Sodium

Primary system Pool Pool Pool

Fuel (UO2)

MOX MOX Metal

Clad material Austenitic stainless

steels Austenitic stainless steels

CN-1515 Austenitic stainless steels

(ODS)

Core outlet temperature(C)

530 540 550

Liner power density(W/cm)

430 430 450

Maximum burn-up(MWd/kg)

60-100 80-120 120-150

Spent fuel storage Primary storage in

vessel and Water pool temporary storage

Primary storage in vessel and Water pool

temporary storage

Primary storage in vessel and Water pool temporary

storage

Safety system Active shutdown system

Passive DHRS

Active and passive shutdown system

Passive DHRS

Active and passive shutdown system

Passive DHRS

10

Basic technical R&D

(1965—1987)

Applied technical

R&D(1987-

1992)

Engineering

technical R&D

(1992—2012)

11

Engaged in the basic theory of SFR and the

principally experiments R&D

Engaged in the applied technical R&D oriented

to the engineering

Engaged in the design and the construction of

CEFR

• The SFR technical R&D started from the 1960’of the last century

• 14 facilities about neutron, thermal-hydraulic, sodium, fuel and material have been constructed.

• 1970.6.29,DF−VI, the first zero power experiment facility get the first criticality

12

• The R&D of SFR supported by the national high technology program “863”

• 61 programs were carried out by the CIAE and other universities, institutes and factories

• More than 20 experiment facilities and loops were constructed

• Some computer code developed

13

• Supported by the “863”program,focus on constructing an experimental sodium fast reactor with 65MW thermal power and 20MW electrical power, it is called CEFR

Site of CEFR:South

west of Beijing city

about 45km

14

Parameter Unit Value Parameter Unit Value Thermal Power MW 65 Primary Circuit

Electric Power, net MW 20 Number of Loops 2

Reactor Core Quantity of Sodium t 260

Height cm 45.0 Flow Rate, total t/h 1328.4

Diameter Equivalent cm 60.0 Number of IHX per Loop

2

Fuel MOX (first loading is

UO2)

Secondary Circuit

Linear Power max. W/cm 430 Number of Loop 2

Neutron Flux n/cm2·s 3.7×1015 Quantity of Sodium t 48.2

Bum-up, first load max.

MWd/t 60000 Flow Rate t/h 986.4

Inlet/outlet Temp. of the Core

℃ 360/530 Tertiary Circuit

Diameter of Main Vessel(outside)

m 8.010 Steam Temperature ℃ 480

Design Life A 30 Steam Pressure MPa 14

Flow Rate t/h 96.2

15

No Content date

1 Project approved 1995.12.29

2 Primary design approved 1997.11.4

3 FCD 2000.5.30

4 Closed of the nuclear island building 2002.8.15

5 Installation finished 2008.12.25

6 Satisfy all the requirement for the first loading 2009.9.27

7 Fist physics criticality 2010.7.21

8 The B stage commissioning work finished 2010.11.30

9 Fit the project target: Connect to the national grid firstly

and 40% rated power operation 24 hours 2011.7.22

10 Restart carry out the power operation experiment 2014.3.14

11 40% power planed experiment finished 2014.5.19

12 Achieve the 100% rated power firstly 2014.12.18

16

2014.3 2014.12

restart,2 test in low power

Loss of vacuum test at 40% level

Load shedding test at 75% level

Over power protection test at

50% level

First time Operation 72h 100% power

17

Overhaul 2016

18

Totally operation 3600hr, produce electricity 16.33 million kwh

• CEFR operated 23days at 39MWt level in 2016.

• The main work to CEFR is overhaul include

– Primary and secondary sodium pump maintains

– Fuel handling machine maintains

– Thermal insulation system upgrade

– Conventional island maintains

– DCS upgrade

– Maintenance and repair of nuclear island auxiliary system

– Instrument and industrial television system repair and rectification

– Radiation monitoring system repair and rectification

– Electrical system repair and rectification

– Safety system and reactor protection system rectification

Water side of SG

19

Parameters Value

Thermal Power, MW 1500

Electricity Power, MW 600

Efficiency 40%

Design load factor 80%

Fuel MOX

Burnup (max),MWd/kg 98

BR 1.15

Circuit Number per loop 2/2

IHX number per circuit 2

CDF <10-6

LERF <10-7

20

1. Try to Fit the GIF Technical Requirement on The Safety and reliability

2. So the off-side emergency could be not necessary from the design

3. D-rap methodology is used in the design of CFR600

4. Ensures good inherent safety.

5. No large positive reactivity insertion at operating conditions or accident conditions.

6. Two independent shutdown system with one passive additional.

7. Decay heat removal system at operating conditions and accident conditions.

8. Primary and secondary Containment design

21

1. “Suspend liquid” passive shutdown system

2. Passive Decay Heat Removal System connected with the hot pool

3. Passive Reactor vessel overpressure protection system

4. “Siphon” device to prevent large amount primary sodium leak after primary pipe break

5. Passive sodium leak stoppage system to mitigation the large water-sodium reactor accident

22

1. System Transient Analysis Code

2. Core and Primary Circuit Thermal-Hydraulic Design code

3. Decay Heat Removal Capability Analysis Code

4. Severe Accident Analysis Code

5. Core Damage Evaluation code

6. Fuel subassembly characteristics evaluation code for the whole operation state

23 Reactor core thermal-

hydraulic analysis Inter subassembly nature

circulation flowrate

Core seismic defamation analysis

System dynamic analysis code

Parameter Value

Thermal power,MW ~2900

Electric power,MW 1200

Thermal efficiency ~41%

Loading factor >85%

Design life,year 60

Fuel MOX ( (TRU,U)O2 )

Cladding ODS

Maximum burn-up,MWd/kg 150

Breading ratio 1.2

Loops per circuit 4/4

CDF <10-6

LERF <10-8

24

25

Main Technical Features of CCFR

1. An innovative pool-type advanced SFR

2. Plant is designed to meet the requirements of the Generation IV nuclear energy systems.

3. More design solutions will be considered, including supercritical CO2 conversion.

4. The technical selection will be consider the continuity with the CDFR

5. More advanced safety design will be considered

CFR1200 diagram based super-critical CO2

Generally Configuration of the Main Heat Transfer System of CFR1200

26

27

2015 2035 2020 2025 2030

Jan2015–Dec2020

Pre-concept Design Jan2021–Dec2024

Concept Design

Jan2025–Dec2028

Preliminary Design Jan2029–Dec2034

Detail Design

Construction

Dec2028

FCD

Dec2020

Decided to build

Dec2034

operation

Suggested Design Schedule for CCFR

28

• R&D status of Gen-IV

29

30

• The Technology Roadmap (2002), defined and planned the necessary R&D and associated

timelines to achieve these goals and allow deployment of Generation IV energy systems after

2030. This roadmapping exercise was a two-year effort by more than 100 international experts

to select the most promising nuclear systems. In 2002, GIF selected the six systems listed below,

from nearly 100 concepts, as Generation IV systems:

• gas-cooled fast reactor (GFR);

• lead-cooled fast reactor (LFR);

• molten salt reactor (MSR);

• sodium-cooled fast reactor (SFR);

• supercritical-water-cooled reactor (SCWR);

• very-high-temperature reactor (VHTR).

31

China SCWR design concept with pressure vessel---CSR1000

Parameters Value

Thermal power 2300MW

Electric power ~1000MWe

Efficiency ~43%

Operating pressure 25MPa

Design pressure 27.5MPa

Reactor inlet temperature 280℃

Reactor outlet temperature 500℃

Reactor flow rate 4284t/h(1190kg/s)

Loop number 2

Cycle direct once through

Coolant flow-path Two-pass

Design lifetime 60 years

SCWRs are high temperature, high-pressure, light water reactors that operate above the thermodynamic critical point of water (374°C, 22.1 MPa). The reactor core may have a thermal or a fast-neutron spectrum, depending on the core design. The concept may be based on current pressure-vessel or on pressure-tube reactors, and thus may

use light water or heavy water as a moderator.

32

SCWR Main Features • High efficiency (up to 48%)

• Simplification of plant components and layout • Design flexibility

R&D and Chief Designer

• Nuclear Power Institute of China

Key Laboratory for nuclear fuel and materials

Key Laboratory for nuclear reactor thermal hydraulic technology

Key Laboratory for nuclear reactor system design

33

• MSR Main Features • MSRs can be divided into two subclasses. In the first subclass, fissile material

is dissolved in the molten fluoride salt.

• In the second subclass, the molten fluoride salt serves as the coolant of a coated particle fuelled core similar to that employed in VHTRs.

• TMSR(Thorium Molten Salt Reactor) project

• Aims : Develop Th-Energy, Non-electric application of Nuclear Energy based on TMSR during coming 20-30 years.

• TMSR-SF(Solid-Fuel), a preliminary design.

• TMSR-LF(Liquid-Fuel), a conceptual design.

34

Power 10 MWt

Lifetime 20 year

Operation time 100 EFPD for single batch of fuel

Average power density 4.0 MW/m3

Fuel element / abundant / 235U load 6cm ball / 17.0% /15.6 kg

Coolant( 1st loop, 2nd loop) FLiBe( 99.99%Li7), FLiNaK

Structure material N alloy, graphite

Reactor coolant inlet temperature 600 ℃

Reactor coolant outlet temperature 650 ℃

Vessel temperature / pressure designed 700C / 0.5MPa (abs.)

Vessel upper cover temperature designed <350℃

1st /2nd loop coolant flow rate 84 kg/s / 150kg/s

Cover gas / pressure Ar / 0.15MPa (abs.)

35

• LFR Main Features • LFRs are Pb or Pb-Bi-alloy-cooled reactors operating at atmospheric pressure and at high

temperature because of the very high boiling point of the coolant (up to 1 743°C).

• The core is characterized by a fast-neutron spectrum due to the scattering properties of lead.

• In China, the Chinese Academy of Sciences (CAS) started in 2011 a new effort to develop an ADS. A new project CiADS(China initiative Accelerator Driven System) is suggested.

• CiADS Design parameters:Beam power250MeV@10mA; Reactor power10MW

36

CiADS Project Owner Chinese Academy of Sciences Guangzhou Branch

Designer Institute of Modern Physics, CAS

partner Institute of High Energy Physics, CAS Hefei Institute of Physical Science, CAS CIAE CGN ……

325 MHz @IHEP Venus-II: Zero-power facility

• Milestones • 2014.06:site identified (Huizhou,

Guangdong Provence)

• 2015.12:Project proved by government

• 2017.04:Feasibility report review

• 2017.09:Preliminary design review

• 2017.10:Construction permission

• 2023.12:Operation

37

• VHRT Main features • The VHTR is a next step in the evolutionary development of high-temperature gas-

cooled reactors.

• It is a graphite-moderated, helium-cooled reactor with thermal neutron spectrum.

• It can supply nuclear heat and electricity over a range of core outlet temperatures between 700 and 950°C, and potentially more than 1 000°C in the future.

• HTR-PM project • High Temperature Gas-cooled

Reactor-Pebble bed Modules

• Designer • Institute of Nuclear and New

Energy Technology(INET) of

Tsinghua University

38

Parameters Design Value

Reactor power, MWt/MWe 250*2/212

Reactor pressure vessel inside diameter, mm 5700

Helium pressure of primary loop, MPa 7.0

Inlet/outlet helium temperature,℃ 250/750

Number of fuel elements in equilibrium core 420,000

Main feed-water temperature,℃ 205

Main steam temperature,℃ 571

Main steam pressure, MPa 13.9

Feed-water flow rate for one reactor steam generator, kg/s 98

Technical features:

‒ Inherent safety

‒ High thermal

efficiency

‒ Pebble-bed modular

‒ Short construction

period

39

40

• China needs a significant nuclear energy capacity in future base on the forecast of the energy totally needed and the environment challenge.

• Nuclear fuel cycle is determined from the perspective of sustainable development of nuclear energy.

• China should develop the closed nuclear fuel cycle based on the fast reactor, and it is should be sustainable, safe, and economic.

• The stratagem of the nuclear energy development is thermal reactor, fast reactor and fusion reactor.

• The stratagem of the fast reactor development is CEFR, CDFR and CCFR. The main tasks of the FR in China are to Raising the utility ration of uranium resource and Transmutation of long life radioactive material. All these function will support a sustainability nuclear energy.

• Several other Gen-IV reactor type are under developing be different Chinese institute or universities.

41

42