SVBR Reactor Plants

SVBR Reactor Plants

SVBR
Reactor Plants
Innovative nuclear technology –
no analogues in the world
Engineering basis of SVBR-type reactors
SVBR-type reactors were designed within the framework of the conversion of unique Russian technology for leadbismuth
coolant marine reactors.
World community shows still
growing interest in nuclear
power deeming it to be one of
the main sources to meet the
increasing need for power generation
for the steady development
of the mankind under the
conditions of gradual depletion
of nonrenewable sources
of energy and environmental
pollution.
A search for reactor technologies that offer the prospect for XXI century is
under way all over the world. The participants in GenIV Forum have chosen
6 innovative reactor systems that incorporate lead-coolant systems, eutectic
lead-bismuth alloy included that meet the Generation IV goals.
Russia possesses unique experience in development and operation of
lead-bismuth coolant submarine reactors.
On the basis of lead-bismuth reactor technology EDO “Gidropress”, IPPE
Russian Research centre and “Atomenergoproyekt” are developing the
designs of small-power SVBR reactors (SVBR stands for the Russian acronym
of lead-bismuth fast reactor) to create nuclear sources of power within the
range from 6 MW-e to 100-400 MW-e, depending on the requirements of
the Customer.
Fig.1. Alfa nuclear submarine
SVBR reactors provide:
Two land prototypes and eight submarines reactor with lead-bismuth coolant have been constructed.
Total operating time of the installations is about 80 reactor-years.
ÔÔ A high level of inherent and passive safety
ÔÔ Considerable simplification of the design of reactor and NPP as a whole in comparison
with the traditional nuclear technologies
A new nuclear power technology that has no analogues in the world was demonstrated on an industrial
scale.
ÔÔ A possibility to work with different kinds of nuclear fuels in different fuel cycles at their
length for not less than 7 years
SVBR reactor performance
ÔÔ Technological support in meeting the non-proliferation requirements
ÔÔ Conservative approach at designing. Orientation to the existing fabrication facilities
and structural materials
ÔÔ Small size and maximum in-shop availability of reactor
ÔÔ Possibility of serial designing of NPP of different power and purpose as well as on-line
methods in construction activities
ÔÔ Competitiveness in the electricity market and NPP attractiveness for capital investments
with a high potential for further improvement of technical and economic indices
The selected power level for the installations, the physical properties of a fast reactor, natural properties of lead-bismuth
coolant and integral design of the reactor make it possible to realize the inherent and passive safety properties
to the utmost, assure meeting the increased requirements for the safety level of GenIV nuclear systems and the
basic criteria of International Project on Innovative Nuclear Reactors and Fuel Cycles.
The combination of properties allows siting the power units with SVBR-type reactors close to population aggregates.
At present SVBR-10 and SVBR-75/100 are the better developed reactor designs (so far conceptual designs have
been worked out).
Main performance of SVBR-10 and SVBR-75/100
Parameter SVBR-10 SVBR-75/100
Reactor thermal power, MW 43,3 280
Functional diversity and area of application
Reactor electric power (gross), MW 12 101,5
Generated steam pressure, MPa 4,2 * 9,5**
Lead-bismuth coolant temperature, inlet/outlet, °С 320 / 480 320 / 482
Power units of a wide power spectrum can be constructed based on the SVBR-type reactor using the
modularity principle to be applied in:
ÔÔ
electric and thermal power generation;
Fuel:
type
average enrichment in U-235,%
UО 2
18,7
UО 2
16,5
Core fuel cycle, thousand, eff.hours 135 53
Time between refuelings, years 15-20 7-8
2
ÔÔ
ÔÔ
sea water desalination;
powered process complexes (for example, coal-chemical, metallurgical and gas-chemical complexes).
* – superheated-410ºC
** – saturated - 307ºC or superheated - 400ºC
3

SVBR
Reactor Plants
SVBR-75/100 -
simple design and increased
safety
SVBR-75/100 reactor meets the most stringent safety requirements (it is human error proof, fail-safe,
proof against sabotage and other ill-intentioned human actions) due to reactor inherent safety resulting
from a combination of reactor type, primary coolant properties and reactor design:
ÔÔ Very high temperature of
lead-bismuth coolant boiling
(~1670ºC) eliminates accidents
due to DNB in the core
and makes it possible to maintain
low primary pressure under
normal operating conditions
and in case of hypothetical accidents;
ÔÔ All primary equipment (Fig. 2)
is housed inside a strong vessel
with a protective housing to
provide an integral (single-unit)
layout. Small free space between
the main vessel and protective
housing prevents loss
of coolant in case the integrity
of reactor main vessel is lost (a
postulated accident);
ÔÔ The level of natural circulation
of the primary and secondary
coolant is sufficient for passive
heat removal under cooldown
conditions;
ÔÔ The cartridge core is located
inside a tank filled with water.
Passive heat transfer via vessel
to the tank water provides passive
cartridge core cooldown
in case all active heat removal
systems fail (a postulated combination
of a number of initiating
events) within at least 5 days
of human non-intervention;
ÔÔ Favorable neutron-physical
properties of lead-bismuth
coolant, a low coefficient of
volumetric expansion in a
combination with the control
algorithms provide an operative
reactivity margin below
1β eff
within the entire fuel cycle,
which in principle excludes a
possibility of reactivity-induced
accidents with prompt neutron
runaway;
ÔÔ Negative reactivity feedbacks
provide power reduction to
the level that does not lead to
core damage in case of an uncontrolled
withdrawal of RCCA
in case of a failure of the highest
worth rod of reactor trip system;
ÔÔ A possibility of chemical explosions
and fires due to internal
causes is ruled out thanks to inherent
safety as the lead-bismuth
coolant remains chemically
inert in case of a loss
of circuit integrity and possible
contact with water and
air. The capability of lead-bismuth
coolant to retain fission
products (iodine, caesium, actinides
- except for inert gases)
can considerably mitigate the
radiological consequences of
a postulated loss-of-coolant
accident;
ÔÔ No materials are applied in
reactor that could evolve hydrogen
either under normal
operating conditions or in accidents;
ÔÔ The assumed primary-to-secondary
pressure ratio, its value
being permanently higher in
the secondary circuit, eliminates
the possibility of radioactive
contamination of steam
to be generated by the system
not only under normal operating
conditions but also in case
of primary-to-secondary leaks
in the SG tubing system;
ÔÔ Low potential energy accumulated
in the primary circuit (low
primary pressure) only restricts
the scale of possible reactor
damage by external impacts.
Protection against external impacts
is ensured by placing the
reactor inside a tight concrete
compartment (Fig. 3);
ÔÔ Due to high safety inherent
to SVBR-75/100 reactor, even
a postulated combination of
such initiating events as concrete
compartment destruction
and a large break of primary
gas system followed by
a direct contact of lead-bismuth
coolant surface with atmospheric
air, does not bring
about reactor runaway, explosion
and fire. Possible radioactive
release is predicted to be
below the level that might require
evacuation of the local
population
Fig. 3.
Layout of reactor equipment
inside a concrete compartment
Modular principle of construction for SVBR-75/100-based NPPs considers the need for a smaller-scale
sources of power for a wide spectrum of potential customers
The modular principle in NPP construction
when a NSSS of a power
Unit is made up of a few SVBR-
75/100 reactor modules jointly operating
for one or several turboplants
makes it possible to construct NPPs
of various power scales depending
on the needs of the customer and
their business solvency (Fig. 4).
SVBR-75/100
4
Fig. 2. Cartridge core reactor
Fig. 4.
Modular NPP with four SVBR-75/100
reactors
5

SVBR
Reactor Plants
Area of possible application for
SVBR-75/100
Nuclear powered water-desalinating facility (NPWDF)
On the basis of standard SVBR-75/100 reactors different-purpose power Units can be constructed. They
can be applied as a part of:
ÔÔ
local different-purpose power sources of different power scale sited in the centers of power consumption;
ÔÔ floating or coastal NPPs to generate electric and thermal power and desalination of sea water both in Russia
and abroad on the basis of the principle: Construction – Ownership – Leasing (or operation). The services can
be rendered to developing nations with the non-proliferation principle considered;
ÔÔ besides, these reactors can be used for renovation to replace the decommissioned power units after their extended
service life has expired. Such a renovation will permit to keep the viability of the NPP satellite towns as
well as the grid, transportation and water infrastructure integrating into the NPP life cycle until the service life of
long-time structures expires.
Layout of co-generating nuclear-powered water desalinating facility:
ÔÔ
ÔÔ
permanent coastal power-generating facility;
replaceable secure transportable autonomous reactor (STAR).
The main design organizations
of the conceptual design
for the NPWDF are:
ÔÔ EDO “GIDROPRESS”,
ÔÔ Russian Research Centre
IPPE,
ÔÔ SPbAEP,
ÔÔ SPMBM «Malakhit»,
ÔÔ Central Research Institute
named after A.N.Krylov.
Local co-generating NPP of average power
6
Rosatom enterprises EDO “Gidropress”,
Russian Research Centre
IPPE, SPbAEP, SPMBM «Malakhit»,
Central Research Institute named
after A.N.Krylov have developed
a conceptual proposal on local
co-generating NPP made up of 4
SVBR-75/100 reactors (Fig. 5). The
prominent features of inherent and
passive safety as well as a relatively
low level of SVBR-75/100 unit power
allow siting the co-generating NPPs
in the vicinity of centers of populations
(at the town development
boundary). At this, the expenditure
on heat transport to the users considerably
decreases.
Technical-economic indices for co-generating NPP
with 4 SVBR-75/100 reactors (in roubles, as of 1991).
Parameter
Co-generating NPP power
- electric, max., MW
- electric, nominal, MW
- heat generation capacity, Gcal/h
Capital investments, thousand rbl.,
Including investments for production of:
- electricity
- thermal power
Specific investments for production of:
- electricity, rbl/kW
- thermal power, thousand rbl/Gcal/h
Net cost of sold-off:
- electricity, kopecks /kW.h
- thermal power, rbl/Gcal
Investment pay-back period
- undiscounted, years
- discounted, years
Value
406
380
520
466 590,82
326 613,57
139 977,25
859,5
269,0
1,35
6,43
8,5
17,5
Fig. 5. Local co-generating NPP
with SVBR-75/100 (general view)
Fig. 6. Co-Generating Nuclear-Powered Water Desalinating Facility based
on SVBR-75/100 STAR:
1 – secure transportable autonomous reactor (STAR); 2 – protective dry dock; 3 – building
for steam-turbine plant; 4 – building for desalinating plant pumps; 5 – desalinating
plant modules; 6 – desalinated water storage tanks; 7 – reactor cooling site for coolant
solidification before transportation; 8 – office building.
Fig. 7. SVBR-75/100 secure transportable autonomous reactor (STAR)
Performance of Co-Generating Nuclear-Powered
Water Desalinating Facility
Parameter
Service life of STAR between refuelings, years 8
Value
Maximum output of fresh water, thousand m 3 /day 200
Electric power of nuclear-powered water desalinating facility
with TG operating in the mode of condensing, MW
80
Power output into grid at maximum fresh water output, MW 9,5
A nuclear powered water desalinating
facility consists of two types
of plants: distillation water-desalinating
plant (DDP) of multi-stage
evaporation and reverse osmosis
water-desalinating plant (RODP).
Installed power of water-desalinating
plants of both types is the same
and amounts to 50 % of the desalinating
facility installed power
which is 100 000 m3/day. Salt content
in DDP desalinated water is 20
mg/l, salt content of RODP desalinated
water is 200 mg/l.
Secure transportable autonomous
reactor (STAR) (Fig. 7) resembles a
replaceable “nuclear storage battery”.
The STAR is supplied based on
the principle: Construction – Ownership
– Leasing for a period determined
by reactor core cycle (at
least 8 years). Supplier runs all the
financial and radiation risks of STAR
construction, transportation, operation
and probable accidents.
7

SVBR
Reactor Plants
Economic indices of NPWDF
For the Customer the cost of construction and operation of nuclear-powered water desalinating facility
amounts to:
ÔÔ capital costs ~ 260 M$, including:
- coastal structures– 60 M$;
- DDP equipment – 120 М$;
- RODP equipment – 80 M$;
ÔÔ annual costs ~ 30 M$/year, including:
- rent for the STAR with account for shipment ~ 12 M$/year;
- cost of NPWDF operation and services ~ 18 M$/year.
The term of NPWDF recoupment and the crediting rate are determined by the Utility for a specific facility
site depending on the local tariffs for fresh water. The accepted rent for the STAR of 12 M$/year will
make it possible:
ÔÔ
ÔÔ
LWR NPP renovation after reactor
service life had expired
Estimated cost of STAR
construction for the Supplier
will be ~ $ 44 million including
the cost of the first fuel
cycle.
for the Supplier to attract investments for STAR construction on the basis of a commercial credit;
3
for the Utility to provide competitive price of the products produced (for example, fresh water ~ 1 $/m and electric
energy ~ 0,035 $/kW*h) and commercial attractiveness of the project (for example, the period of project recoupment
is ~12 years at ~ 10% crediting rate for the loaned capital.
Renovation of NPP Units with light-water reactors after reactor service life has expired is realized by siting the required
number of SVBR 75/100 reactors in the emptied compartments that used to house steam generators and reactor
coolant pumps to create a modular structure of the renovated unit.
SVBR-10 – reactor with
super-long fuel cycle for
local power supply
On the basis of SVBR-10 reactors power units of various purpose and
power range can be constructed without on-site refueling in coastsited,
floating and land-based embodiments.
These power units can be used to generate electric and thermal
energy, sea water desalination etc.
The main organizations to
develop SVBR-10 reactor
conceptual design are:
ÔÔ EDO “Gidropress”,
ÔÔ Russian Research Centre
IPPE.
Small transportable autonomous reactors for small-power
coastal NPPs
The concept of costal small-power NPPs was developed in 2005 that incorporate coastal structures and replaceable
STARs with SVBR-10 to be delivered to the NPP site and taken back to the Supplier country by sea for core
refueling (Fig. 9).
Fig. 9. Coastal small-power NPP
based on SVBR-10 STAR
8
Fig. 8. Renovation diagram for a NPP with a light water reactor
The results of technical and economic feasibility studies for NV NPP Units 2, 3 and 4 renovation on the basis of SVBR-
75 have shown that the construction of new power Units to replace the decommissioned ones is twice as expensive
as the renovation as far as the specific investments are concerned.
STAR (Fig. 10) is 8 m in diameter and 11,2 m high, its weight being 310 t with
coolant. Cartridge reactor (without the core) is completely fabricated at
the manufacturing plant.
The coast-sited NPP is a preferable option for the Customer that has already
developed industrial or social infrastructure or intends to develop it. Otherwise,
the floating-type NPP will be a better option for the Customer.
Fig. 10. SVBR-10 secure transportable
autonomous reactor
9

SVBR
Reactor Plants
Floating NPP with two SVBR-10 reactors
Conceptual design of a floating-type NPP with two SVBR-10 (Fig. 11) reactors was developed by JSC “Atomenergo”
and reactor design organizations. The results of design development state that a possibility exists to create a power
source that possesses a higher safety level and a better economic efficiency in comparison with a floating NPP with
water-cooled and water-moderated reactors.
Avenues of further development
SVBR-type reactor designs are aimed at providing operation with different types of fuel and in different fuel cycles
that most correspond to each stage of nuclear power engineering development without changing the construction
or safety characteristics deterioration. At the first stage the usage of proven uranium oxide fuel and operation
in an open fuel cycle with a delayed recycling is offered. In a closed fuel cycle with MOX-fuel applied, the reactor
will operate in the mode of fuel regeneration. When nitride fuel is applied fuel breeding will be provided. Spent fuel
from VVER and RBMK reactors can be directly used as the make-up fuel without separating uranium, fission products,
plutonium and minor actinides.
After appropriate experimental study has been performed, an increase in reactor parameters and, consequently,
unit power is possible.
Status of work in lead-bismuth reactor
technology
Fig. 11. Floating NPP with two SVBR-10 reactors
The application of a fast reactor lets decreasing the amount of cartridge cores used within the period
of operation and refrain from on-site refuelings which permits to eliminate additional refueling facilities
and fresh and spent fuel storage structures in the design of the floating NPP.
Besides, the well-developed inherent and passive safety features lead to simplification of the NPP reactor and subsequently,
to a reduction in the amount of its systems and equipment.
Project economic efficiency
To estimate the economic efficiency
of the designed floating
NPP the Central power park of the
Kamchatka Region was chosen to
be the area of its siting. The selling
rates for heating and electricity
are accepted corresponding to
the average tariffs in the Central
power park as of 01.01.2006.
Integral efficiency indices for floating NPP with SVBR-10 reactor
Index
SVBR-10
The economic efficiency of the floating NPP construction project was calculated in accordance with the methodological
recommendations approved by the Ministry for Economics and Gosstroy of Russia.
Floating NPP installed power, MW-e 24
Simple recoupment period, years 10
Discounted recoupment period (d=8%), years 16
Value
Internal return on assets 10,5%
Profitability index 1,23
The calculations assumed Project financing by bank crediting at 8 % annual rate in currency. The credits are loaned
in three tranches with a 4-year term of each tranche payback. The calculations show that the payables will be paid
off 8 years after the first tranche has been transferred.
The results of the work in the given direction were considered
at the meeting of the Scientific and Technical Council
of Rosatom of Russia and the decision reads:
“The scope and extension of SVBR-75/100 design development,
maximum possible usage of proven engineering decisions, available
structural materials and fuel infrastructure and the capacities
of the machine-building industry as well as the available
achievement in research and development to substantiate the
lead-bismuth technology lead to a conclusion that the reactor
technology can be realized in a pilot power plant project to prove
the possibility and efficiency of the promising application of the
technology”.
It was found expedient to go on designing a prototype power unit
with SVBR-75/100 reactor for a selected site.
The task to construct the prototype power unit with SVBR-75/100
reactor is part of the Federal Target Programme “Development of
nuclear power complex of Russia in 2007-2010 and directions up
to 2015”. At present the development of detailed project report is
under way.
SVBR-75/100 project was the
winner at the Innovation Forum
II of Rosatom in 2007 in the
nomination “Nuclear power industry”.
The calculations show that with the average selling rate of thermal energy of 40 $/Gcal, the minimum (boundary)
selling tariff for electricity will be 4,8 cents/kW-h. Specific process costs (tax-free) will amount to 1,5 cents/kW-h with
account for the kilowatt-hours in the form of heat supplied to consumers.
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Processability indices
All equipment for the SVBR reactor-based power Units can be manufactured at the machine-building plants of the
Russian Federation. Since no unique machine-building equipment is required to manufacture the reactor (in comparison
with vessels for water-cooled and water-moderated power reactors), a possibility for competitive manufacturing
arises. Advanced methods of serial designing and on-line methods in construction activities can be used at
modular-type power units construction which considerably shortens the terms and cost of work performance.
11

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