MayMay 2010 2010

FEDERAL SCIENTIFIC PRODUCTION CENTERJSC “Afrikantov OKBM”

EFFECTIVENESS EVALUATION FOR THE FAST SODIUM-

COOLED REACTOR DESIGN SOLUTIONS AND THEIR EVOLUTION IN

NEW DESIGNS

B.A. Vasilyev

7th International Scientific and Technical Conference "Nuclear Power Safety, Effectiveness and Economy", MNTK-2010

BN FAST REACTOR TECHNOLOGY STATUS

In February 2010, the Government of the Russian Federation approved the Federal Target Program “New Generation Nuclear Power Technologies for the Period 2010-2015 and for the Future to 2020” (FTP NGNT)

As part of the FTP NGNT, R&D work is provided for the BN-1200, next generation sodium-cooled fast reactor with the electric power of 1200 MW

BN-1200 is being developed such that the task be resolved for serial construction of the BN-1200 reactors after 2020

The BN-800 reactor is under construction; the planned completion date for the construction is 2013

On April 8, it was 30 years since the BN-600 reactor commissioning. The BN-600 is the only operating fast power reactor in the world

PROVEN DESIGN SOLUTIONS. BN-600 DESIGN

Activity on fast reactors in Russia was started in 1960 by designing the first pilot industrial BN-350 power reactor. The reactor was commissioned in 1973 and was in operation until 1998

In 1980, the next, more powerful BN-600 reactor was commissioned in Beloyarsk NPP

In April 2010, the reactor completed its assigned service life of 30 years.

Its service life extension to 45 years has been validated

BN-600 30-YEAR OPERATION EXPERIENCE

Capacity factor 78% over the last 5 years (close to the capacity factor of serial VVERs equal to 79.9% over the same period)

Reactor scrams The average number of reactor scrams for 7000 operating hours is 0.2 (for NPPs across the world, it is 0.5–0.7). No reactor scram was in the period 2000 through 2009

Average radioactive gas release over the last 6 years

1% of the allowable level (4 times lower as compared to VVER NPPs over the same period)

Collective personnel dose rate over the last 5 years

0.54 man·Sv per year (2.2 times lower than the same for VVER NPPs)

SODIUM LEAKS IN BN-600

There have been 27 outside leaks (5 of them were radioactive sodium leaks) and 12 SG leaks. The main reason for them is deviations in the manufacture quality of auxiliary pipelines

An only radioactive sodium leak (~1 m3) resulted in a radioactive substance release into atmosphere below the allowable limits for normal operation of the NPP

The last sodium leak took place in the BN-600 reactor in 1994

Over the last 24 operational years, only one small leak took place in the SG

Capacity factor reduction due to the leaks is negligibly small

The reliability of design measures to prevent and localize inter-circuit and outer sodium leaks has been convincingly demonstrated

BN-600 CAPACITY FACTOR VARIATIONКИ

УМ

,%

56,51

71,76 72,75 72,48 73,46 74,1176,6 75,89

65,91

69,83

83,5380,29

78,19

70,31

76,3272,97

47,93

76,4373,23

79,8977,35

75,74

80,0477,75 78,6 77,78 77,49 76,53

0

10

20

30

40

50

60

70

80

90

100

1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 2007 2008 2009

Cap

acity

, %

BN-600 REACTOR CORE MODIFICATIONS

Characteristics  Modification

01 01M2

Operation period for the core option 1980-1986 from 2005

Core height, mm 750 1030

Axial blankets height, mm

- upper

- lower

 

400

400

 300

350

Number of fuel enrichment zones 2    3

Core structural materials:

- Fuel cladding

- FA shroud

EI-847

Cr16Ni11Mo3

ChS-68cw

EP-450

Maximum fuel rod linear heat rating, kW/m  54.0 48.0 

Maximum fuel burnup, % h.a. 7.2 11.1

Maximum core life, eff. days (central/peripheral FAs) 200/300 560/720*

Average fuel burnup, MW·d/kg U 42.5 70.0

Maximum damage dose, dpa 43.5 82.0

TASKS TO BE SOLVED BY THE BN-800 CONSTRUCTION AND OPERATION

Breeding mode of operation with MOX-fuel

Pilot-scale demonstration of key closed fuel cycle components

Developing innovative technologies for future LMFBR:

advanced fuel and structural materials testing and qualification

MA burn-out technology demonstration

testing novel technical solutions sustaining competency in the

LMFBR technology

BN-800 – important milestone in evolutionto next generation nuclear power technology

THE BN-800 UNIT DESIGN EVOLUTION AND KEY FEATURES

1984 – initial design– evolutionary up-rated version of the BN-600 1994 – updated design approval

Reactor plants data BN-600 BN-800 Reactor thermal capacity, MW 1470 2100 Power unit electric capacity, MW 600 880 Nuclear fuel UO2 MOX Sodium coolant temperature

- core inlet 377 354 - core outlet (IHX entry)

550 547

Superheated steam - temperature, °C 505 490 - pressure, MPa 14.2 13.7

Intermediate superheating media sodium steam Reactor vessel dimensions, m

- diameter 12.86 12.9 - height 14.7 15.0

Specific metal intensity, t/MWe 13.0 9.7

Distinctive design features:

• much higher unit capacity• passive safety systems• MOX-fuel

Additional passive emergency protection system is provided for

Emergency decay heat removal system is introduced that utilizes air heat exchangers

A tray is provided for localization of molten core debris in a postulated accident with a failure of all reactor protection equipment

IMPROVEMENTS IN BN-800 DESIGN SOLUTIONS(1)

Introduce technical solutions to enhance safety, efficiency and reliability of the power unit:

IMPROVEMENTS IN BN-800 DESIGN SOLUTIONS (2)

Hand operations are eliminated from the refueling system to ensure the possibility to handle fresh FAs with highly radioactive MOX-fuel

The design lifetime is increased from 30 years (BN-600) to 45 years with the possibility of future extension to 60 years

MOX-fuel burnup is increased through the replacement of ChS-68 austenitic steel (burnup is up to 10% h.a.) by EK-164 c.w. (up to 13% h.a.); and then by ferritic-martensitic steel (up to 15% h.a.)

VIEW OF THE BN-800 CONSTRUCTION SITE. MAY 2010

Support belt

Vessel bottom

EVOLUTION OF DESIGN SOLUTIONS IN BN-1200 (1)

The primary circuit layout is integral with the safety vessel and lower vessel support

The rotating plugs of the in-reactor refueling system have sealing hydraulic locks based on tin-bismuth alloy

Separate suction cavities for the primary circuit pumps with check valves in the discharge nozzles make it possible to isolate one of the three primary loops without a reactor shutdown in case of equipment failure

There is an in-reactor spentFA storage

Main design solutions, which proved to be successful in BN-600 and used in BN-800:

CPS column

Intermediate heat

exchanger (IHX)

Pressure chamber

CoreSupport belt

Tray

Pressure pipeline

MCP-1 Refueling mechanism

Rotary plugs

Safety vessel

Reactor vessel

ECS AHE

Improved reactor and SG designs (reduced materials consumption)

Bellows-type compensators in the secondary circuit

pipelines (reduced length and material consumption) Substantially simplyfied refueling system as compared to BN-600 and BN-800 (reduced materials consumption)

New Solutions:

EVOLUTION OF DESIGN SOLUTIONS IN BN-1200 (2)

The emergency heat removal system uses autonomous heat exchangers in-built into the reactor vessel (enhanced reliability) Primary sodium cold traps are located in the reactor vessel (pipelines

containing radioactive sodium and their auxiliary systems are eliminated)

Бак аварийногосброса

Парогенератор

Промежуточныйтеплообменник

Бак аварийногосброса

Воздушныйтеплообменник

Автономныйтеплообменник

Реактор

Буфернаяемкость

Главныйциркуляционныйнасос II контура

Ограждениегерметичное

надреакторногообъема

Бакрасширительный

Бак аварийногосброса

Парогенератор

Промежуточныйтеплообменник

Бак аварийногосброса

Воздушныйтеплообменник

Автономныйтеплообменник

Реактор

Буфернаяемкость

Главныйциркуляционныйнасос II контура

Ограждениегерметичное

надреакторногообъема

Бакрасширительный

Бак аварийногосброса

Парогенератор

Промежуточныйтеплообменник

Бак аварийногосброса

Воздушныйтеплообменник

Автономныйтеплообменник

Реактор

Буфернаяемкость

Главныйциркуляционныйнасос II контура

Ограждениегерметичное

надреакторногообъема

Бакрасширительный

SecondaryMCP

Buffer tank

Emergency dump tank

Steam generator

Intermediate heat exchanger

Emergency dump tank

Leak-tight cover for above-the-reactor space

Expansion tank

Air heat exchanger

Autonomous heat exchanger

Reactor

BN-1200 RP Layout

 Characteristics BN-1200

Nominal thermal power, MW 2900

Electric power, gross, MW 1220

Number of heat removal loops 4

Primary coolant temperature, C:

- IHX inlet/outlet

 

550/ 410

Secondary coolant temperature, C:- SG inlet/outlet 527/355

Third circuit parameters:

- live steam temperature, C- live steam pressure, MPa

- feedwater temperature, C

510

14

240

BN-1200 DESIGN DATA

DESIGN OF THE MAIN EQUIPMENT

Technical solutions for MCP-1, MCP-2, IHX in the BN-800 and BN-1200 are basically the same as those in the BN-600

The BN-800 CRDM design was upgraded through simplification of the kinematic chain and enhancement of its reliability; the specific metal intensity was reduced. A similar technical solution will be

used in the BN-1200

The BN-800 steam generator design is characteristic of the less number of modules (20 per loop instead of 24 per loop utilized in the BN-600) due to elimination of sodium intermediate steam

superheaters. The BN-1200 SGs are substantially enlarged: 2-4 moduls per loop.

Technical solutions for the AHX are the same as those in the BN-800 – finned tubes, heat removal by air natural

convection.

Enlarged fuel rods ( 6.9 mm 9.3 mm, reduced average fuel heat rating, increased FA life)

Enlarged FA (S=96 mm S=181 mm, reduced number of FAs)

Increased fuel volume fraction (0.43 0.47, increased breeding factor)

Increased gas cavity in a fuel rod, Tclad < 670 °C (to ensure high fuel burnup)

Use of a single fuel enrichment zone instead of three ones (simplified fuel production)

In-reactor storage area that ensures 2-year fuel delay(simplified refueling)

NEW SOLUTIONS FOR THE CORE BN-1200

Parameter Value

FA core residence, eff. day 1320→1650→1980

Maximum fuel burnup, % h.a. 14.3→17.8→21

Average fuel burnup for unloaded FAs, MWday/kg 93→116→138

Maximum damage dose per FA, dpa 140→170→200

Maximum fuel rod linear power, kW/m 46.5

Breeding factor ~1.2

The reactor core design is being developed to ensure possible transitionto mixed nitride fuel (breeding factor is up to 1.45)

CORE SPECIFICATIONS

REDUCTION IN SPECIFIC METAL INTENSITY FOR THE BN REACTORS

Parameter BN-800 BN-1200

1 Specific RP materials consumption, t/MW(e), including: 9.7 5.6

2 Reactor 3.64 3.47

3 Steam generators 3.42 0.66

4 External fuel handling system 1.16 0.38

5 Other equipment 1.48 1.09

Parameter BN-600 BN-800 BN-1200

Specific RP materials consumption, t/MW(e) 13.0 9.7  5.6

Refueling interval, day 110…170 155 330

Capacity factor 0.77 – 0.8 0.85 0.9 

Lifetime, years 45 45 60

BN-1200 economic performance will be comparable with thatof VVERs having the same power. In perspective, the cost of electricity

generated by BN-1200 should be lower than that of VVER due to expected growth in natural uranium prices.

COMPARISON OF BN REACTORS TECHNICALAND ECONOMIC PERFORMANCE

Properties and technical solutions to ensure safety BN-600 BN-800 BN-1200

1 Main properties:- low pressure- low corrosion activity- high boiling temperature

+

2 Technical solutions

2.1 Emergency protection:- active- active + passive based on hydraulically suspended rods - active + passive based on hydraulically suspended rods + passive temperature-actuated system

+--

++-

+++

2.2 Emergency heat removal system:- belongs to the third circuit- air heat exchanger in the second circuit- air heat exchanger connected to the primary circuit

+- -

 + - +

2.3 Molten fuel tray system - + +

2.4 Emergency release localization system - - +

EVOLUTION OF SOLUTIONS FOR BN REACTOR SAFETY (1)

Thanks to the solutions adopted in the BN-1200 design, safety parameters are planned to be significantly improved:

EVOLUTION OF SOLUTIONSFOR BN REACTOR SAFETY (2)

Core severe damage probability is by an order of magnitude lower than that required by regulatory documents

The exclusion area is within the NPP site for any design accidents

A target criteria has been established: the area for protective measures planning shall coincide with the NPP site boundary for severe beyond-design basis accidents of which occurrence does not exceed 10-7 reactor/year

Experience gained in development and operation of fast sodium-cooled reactors demonstrates effectiveness of basic design solutions adopted in BN-600 reactor, their reliable operation and high safety level

The basic design solutions evolved in the BN-800 and BN-1200 designs. The new design solutions for the BN-1200 will have to be tested by analytical and experimental investigations

The BN-1200 design may be related to Generation IV NPPs due to:

optimal combination of reference and innovative solutions

enhanced safety characteristics

high technical and economic performance

possibility of extensive fuel breeding

CONCLUSIONS