An overview of future sustainable nuclear power reactors

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International Journal of Energy and Environment (IJEE), Volume 4, Issue 5, 2013, pp.743-776
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747
A modern BWR fuel assembly comprises 74 to 100 fuel rods, and there are up to approximately 800
assemblies in a reactor core, holding up to approximately 140t of uranium. The secondary control system
involves restricting water flow through the core so that steam in the top part means moderation is
reduced [8].

3.3 Pressurized heavy water reactor
The pressurized heavy water reactor (PHWR) or CANDU reactor design has been developed since the
1950s in Canada. The acronym CANDU stands for Canada deuterium uranium. All current power
reactors in Canada are of the CANDU type. It uses natural uranium (0.7%
235
U) oxide as fuel, hence
needs a more efficient moderator, such as, D
2
O [3, 9].
The coolant is kept under high pressure to raise its boiling point and avoid significant steam formation in
the core. The hot D
2
O generated in this primary cooling loop is passed into a heat exchanger heating light
water in the less-pressurized secondary cooling loop. The generated steam drives a conventional turbine
with a generator for power generation [9].
The moderator is in a large tank called a calandria, penetrated by several hundred horizontal pressure
tubes which form channels for the fuel, cooled by a flow of D
2
O under high pressure in the primary
cooling circuit, reaching 290°C. Traditional designs using light water as a moderator will absorb too
many neutrons to allow a chain reaction to occur in natural uranium due to the low density of active
nuclei. D
2
O absorbs fewer neutrons than light water, allowing a high neutron economy that can sustain a
chain reaction even in unenriched fuel. Also, the low temperature of the moderator (below the boiling
point of water) reduces changes in the neutrons' speeds from collisions with the moving particles of the
moderator (neutron scattering). The neutrons therefore are easier to keep near the optimum speed to
cause fissioning, therefore, they have good spectral purity. At the same time, they are still somewhat
scattered, giving an efficient range of neutron energies [3, 4].
The large thermal mass of the moderator provides a significant heat sink that acts as an additional safety
feature. If a fuel assembly were to overheat and deform within its fuel channel, the resulting change of
geometry permits high heat transfer to the cool moderator, thus preventing the breach of the fuel channel,
and the possibility of a meltdown. Furthermore, because of the use of natural uranium as fuel, this reactor
cannot sustain a chain reaction if its original fuel channel geometry is altered in any significant manner.
The central functionality behind the CANDU design is D
2
O moderation and on-line refueling, which
permits a range of fuel types to be used, including natural uranium, enriched uranium, thorium, and used
fuel from light water reactors (LWRs). Significant fuel cost savings can be realized if the uranium does
not have to be enriched, but simply formed into ceramic natural UO
2
fuel. This saves not only on the
construction of an enrichment plant, but also on the costs of processing the fuel. However, some of this
potential savings is offset by the initial, one time cost of the D
2
O. The D
2
O required must be more than
99.75% pure and tones of this are required to fill the calandria and the heat transfer system [9].
CANDU reactors do have some drawbacks. D
2
O generally costs hundreds of dollars per kilogram,
though this is a trade-off against reduced fuel costs. It is also notable that the reduced energy content of
natural uranium as compared to enriched uranium necessitates more frequent replacement of fuel, which
is normally accomplished by use of an on-power refueling system. The increased rate of fuel movement
through the reactor also results in higher volumes of spent fuel than in reactors employing enriched
uranium. However, as the unenriched fuel was less reactive, the heat generated is less, allowing the spent
fuel to be stored much more compactly [10].

3.4 Graphite moderated reactors
Gas cooled reactors (GCR) and advanced gas cooled reactors (AGR) use carbon dioxide (CO
2
) as the
coolant to carry the heat to the turbine, and graphite as the moderator. Like D
2
O, a graphite moderator
allows natural uranium, usually in GCR or slightly enriched uranium, usually in AGR, to be used as fuel
[3, 4].

3.4.1 Advanced gas cooled reactor
The advanced gas cooled reactor (AGR) reactor is a British design generation II GCR, using graphite
moderator and CO
2
as coolant. The mean temperature of the hot coolant leaving the reactor core was
designed to be 648°C. In order to obtain these high temperatures, yet ensure useful graphite core life
(graphite oxidises readily in at high temperature) a re-entrant flow of coolant at the lower boiler outlet
temperature of 278°C is utilised to cool the graphite, ensuring that the graphite core temperatures do not
International Journal of Energy and Environment (IJEE), Volume 4, Issue 5, 2013, pp.743-776
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748
vary too much. The superheater outlet temperature and pressure are designed to be 170bar and 543°C.
The fuel is UO
2
pellets, enriched to 2.5-3.5%, in stainless steel tubes. The original design concept of the
AGR was to use a beryllium based cladding. When this proved unsuitable, the enrichment level of the
fuel was raised to allow for the higher neutron capture losses of stainless steel cladding. This
significantly increased the cost of the power produced by an AGR. The CO
2
circulates through the core,
reaching 650°C and then past steam generator tubes outside it, but still inside the concrete and steel
reactor pressure vessel. Control rods penetrate the moderator and a secondary shutdown system involves
injecting nitrogen to the coolant.
The AGR was designed to have a high thermal efficiency of about 41%, which is better than modern
PWRs which have a typical thermal efficiency of 34%. This is due to the higher coolant outlet
temperature of about 640°C practical with gas cooling, compared to about 325°C for PWRs. However
the reactor core has to be larger for the same power output, and the fuel burn-up ratio at discharge is
lower so the fuel is used less efficiently, countering the thermal efficiency advantage. AGRs are designed
to be refueled without being shut down first. This on-load refueling is an important part of the economic
case for choosing the AGR over other reactor types [11].

3.4.2 Water cooled light water graphite moderated reactor
The light water graphite moderated reactor (RBMK) is a Soviet design, developed from plutonium
production reactors. It employs long vertical pressure tubes running through graphite moderator, and is
cooled by water, which is allowed to boil in the core at 290°C, much as in a BWR. Fuel is low-enriched
UO
2
made up into fuel assemblies 3.5m long. With moderation largely due to the fixed graphite, excess
boiling simply reduces the cooling and neutron absorption without inhibiting the fission reaction and a
positive feedback problem can arise [3, 4].
It is estimated that about 5.5% of the core thermal power is in the form of graphite heat. About 80-85%
of this heat is removed by the fuel rod coolant channels, via the graphite rings. The rest of the heat is
removed by the control rod channel coolant. The gas circulating in the reactor plays the role of assisting
the heat transfer to the coolant channels. There are 1661 fuel channels and 211 control rod channels in
the reactor core. The fuel assembly is suspended in the fuel channel on a bracket, with a seal plug. The
seal plug has a simple design, to facilitate its removal and installation by the remotely controlled
refueling machine. The fuel channels may, instead of fuel, contain fixed neutron absorbers or be empty
and just filled with the cooling water. The small clearance between the pressure channel and the graphite
block makes the graphite core susceptible to damage. If the pressure channel deforms, e.g., by too high
internal pressure, the deformation or rupture can cause significant pressure loads to the graphite blocks
and lead to their damage, and possibly propagate to neighboring channels.
The fuel pellets are made of UO
2
powder sintered with a suitable binder into barrels. The material may
contain added europium oxide as a burnable nuclear poison to lower the reactivity differences between a
new and partially spent fuel assembly. To reduce thermal expansion issues and interaction with the
cladding, the pellets have hemispherical indentations. The enrichment level is 2% (0.4% for the end
pellets of the assemblies). Maximum allowable temperature of the fuel pellet is 2100°C. The rods are
filled with helium at 5bar and hermetically sealed. Retaining rings help to seat the pellets in the center of
the tube and facilitate heat transfer from the pellet to the tube. The pellets are axially held in place by a
spring. Each rod contains 3.5kg of fuel pellets. The fuel rods are 3.64m long, with 3.4m of that being the
active length. The maximum allowed temperature of a fuel rod is 600°C. The fuel assemblies consist of
two sets of 18 fuel rods. The rods are arranged along the central carrier rod and held in place with 10
stainless steel spacers separated by 360mm distance. The two sub-assemblies are joined with a cylinder
at the center of the assembly and during the operation of the reactor, this dead space without fuel lowers
the neutron flux in the central plane of the reactor [12].

3.5 Fast breeder reactors
As of 2006, all large-scale fast breeder reactor (FBR) power stations have been liquid metal fast breeder
reactors (LMFBR) cooled by liquid sodium. These have been of one of two designs (a) Loop type, in
which the primary coolant is circulated through primary heat exchangers outside the reactor tank, but
inside the biological shield due to radioactive sodium-24 (
24
Na) in the primary coolant and (b) Pool type,
in which the primary heat exchangers and pumps are immersed in the reactor tank.
All current FBR designs use liquid metal as the primary coolant, to transfer heat from the core to steam
used to power the electricity generating turbines. FBRs have been built cooled by liquid metals other
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749
than sodium (some early FBRs used mercury), other experimental reactors have used a sodium-
potassium alloy. Both have the advantage that they are liquids at room temperature, which is convenient
for experimental rigs but less important for pilot or full scale power stations. Lead and lead-bismuth alloy
have also been used. FBRs usually use a mixed oxide fuel core of up to 20% plutonium dioxide (PuO
2
)
and at least 80% UO
2
. Another fuel option is metal alloys, typically a blend of uranium, plutonium, and
zirconium (used because it is transparent to neutrons). Enriched uranium can also be used on its own.
In many designs, the core is surrounded in a blanket of tubes containing non-fissile uranium-238 (
238
U)
which, by capturing fast neutrons from the reaction in the core, is converted to fissile
239
Pu (as is some of
the uranium in the core), which is then reprocessed and used as nuclear fuel. Other FBR designs rely on
the geometry of the fuel itself (which also contains
238
U), arranged to attain sufficient fast neutron
capture. The
239
Pu (or the fissile
235
U) fission cross-section is much smaller in a fast spectrum than in a
thermal spectrum, as is the ratio between the
239
Pu /
235
U fission cross-section and the
238
U absorption
cross-section. This increases the concentration of the
239
Pu /
235
U needed to sustain a chain reaction, as
well as the ratio of breeding to fission. On the other hand, a fast reactor needs no moderator to slow
down the neutrons at all, taking advantage of the fast neutrons producing a greater number of neutrons
per fission than slow neutrons. For this reason ordinary liquid water, being a moderator as well as a
neutron absorber is an undesirable primary coolant for fast reactors. Because large amounts of water in
the core are required to cool the reactor, the yield of neutrons and therefore breeding of
239
Pu are strongly
affected. Theoretical work has been done on reduced moderation water reactors, which may have a
sufficiently fast spectrum to provide a breeding ratio slightly over 1. This would likely result in an
unacceptable power derating and high costs in an liquid water cooled reactor, but the supercritical water
coolant of the supercritical water reactor (SCWR) has sufficient heat capacity to allow adequate cooling
with less water, making a fast-spectrum water cooled reactor a practical possibility. In addition, a D
2
O
moderated thermal breeder reactor, using thorium to produce uranium-233 (
233
U), is also possible [13].

3.6 Aqueous homogeneous reactor
Aqueous homogeneous reactor (AHR) is a type of nuclear reactor in which soluble nuclear salts, which
are usually uranium sulfate or uranium nitrate, are dissolved in water. The fuel is mixed with the coolant
and the moderator, thus the name homogeneous. The water can be either D
2
O or light water, both which
need to be very pure. A D
2
O AHR can achieve criticality (turn-on) with natural uranium dissolved as
uranium sulfate. Thus, no enriched uranium is needed for this reactor. The D
2
O versions have the lowest
specific fuel requirements (least amount of nuclear fuel is required to start them). Even in light water
versions less than 0.454kg of
239
Pu or
233
U is needed for operation. Neutron economy in the D
2
O versions
is the highest of all reactor designs.
Their self-controlling features and ability to handle very large increases in reactivity make them unique
among reactors, and possibly safest. AHRs were sometimes called water boilers, although they are not
boiling water reactors. They seem to be boiling their water, but in fact this bubbling is from the
production of hydrogen and oxygen as the radiation, and especially the fission particles, dissociate the
water into its constituent gases. Corrosion problems associated with sulfate base solutions limited their
application as breeders of
233
U fuels from thorium. Current designs use nitric acid base solutions (e.g.,
uranyl nitrate) eliminating most of these problems in stainless steels [14].

4. Generation III reactors
Generation III reactors have emerged through the ‘90’s, with evolutionary designs, they are the evolution
of generation II, as illustrated in Figure 1, with significant advances in terms of safety and economics
resulting in near-term deployment in several countries. Some are evolutionary from the generation II
PWR, BWR and CANDU designs, and some designs are more radical. The former include the advanced
boiling water reactor (ABWR), two of which are now operating with others under construction. The best-
known radical new design is the pebble bed modular reactor (PBMR), which uses helium as coolant at
very high temperature to drive a turbine directly. Generation III reactors are undergoing deployment and
will be doing so up to the arrival of generation IV reactors after 2030. Table 2 tabulates the various
generation III reactors designs found in the literature and Table 3 provides the associated capital cost
estimates based on various projects around the world. Generation III reactors have (a) a standardized
design for each type to expedite licensing, reduce capital cost and reduce construction time, (b) a simpler
and more rugged design, making them easier to operate and less vulnerable to operational upsets, (c)
higher availability and longer operating life, typically 60 years, (d) reduced possibility of core melt
International Journal of Energy and Environment (IJEE), Volume 4, Issue 5, 2013, pp.743-776
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750
accidents, (e) minimal effect on the environment, (f) higher burn-up to reduce fuel use and the amount of
waste and (g) burnable absorbers to extend fuel life. The greatest departure from generation II designs
are the passive or inherent safety features that require no active controls or operational intervention to
avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to
high temperatures [15].



Figure 1. Nuclear reactors evolution

Table 2. Generation III reactors designs

No. Reactor Capacity (MWe) Power cycle
ALWR Advanced light water reactors
1 EPR European pressurized water reactor 1600-1750 Rankine
2 ABWR Hitachi 600-1700 Rankine
3 ESBWR Economic simplified boiling water reactor 1390-1550 Rankine
4 APWR Advanced pressurized water reactor 1500 Rankine
5 BWR 90+ 1500 Rankine
6 VVER-448 1500 Rankine
7 APR - 1400 Advanced pressurized water reactor 1400
(System 80+)
1400 Rankine
8 ABWR Advanced boiling water reactor 1300 Rankine
9 SWR-1000 Siedewasser boiling water reactor 1000-1290 Rankine
10 AP1000 Advanced passive 1000 1100 Rankine
11 VVER-91 1000 Rankine
12 V-392 950 Rankine
13 VVER-640 640 Rankine
14 VPBER-600 600 Rankine
15 AP600 Advanced passive 600 600 Rankine
16 IRIS International reactor innovative and secure 335 Rankine
17 MSBWR Modular simplified boiling water reactor
(under development)
50 & 200 Rankine
18 IRIS-50 International reactor innovative and secure
(under development - GIII+)
>50 Rankine
19 KLT-40 30-35 Rankine
20 TRIGA power system (pressurized water reactor) 16,4 Rankine
21 VBER-150 110 Rankine
22 VBER-300 295 Rankine
23 VK-300 (under development - boiling water reactor) 250 Rankine
24 ABV (under development - pressurized water reactor) 10-12 Rankine

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751
Table 2. (Continued)

No. Reactor Capacity (MWe) Power cycle
25 CAREM (under development) 27 Rankine
26 SMART system - integrated modular advanced
reactor
110 Rankine
27 MRX (under development) 30 Rankine
28 NP-300 100-300 Rankine
29 NHR-200 N/A Rankine
PHWR's Pressurized heavy water reactors
30 CANDU-9 Canadian deuterium uranium 925-1300 Rankine
31 ACR-1000 Advanced CANDU reactor 1000 hybrid
PHWR/PWR
1100-1200 Rankine
32 CANDU-X 350-1150 Supercritical
Rankine
33 AHWR Advanced heavy water reactor 300 Rankine
34 ACR-700 Advanced CANDU reactor 700 hybrid
PHWR/PWR
750 Rankine
HT GCR's High temperature gas cooled reactors
35 GTHTR Gas turbine high temperature reactor 300 Brayton
36 GT- MHR Gas turbine modular helium reactor 285 Brayton
37 HTR-PM High temperature pebble bed gas cooled
reactor
195 Rankine
38 PBMR Pebble bed modular reactor 165 Brayton
39 HTTR High temperature test reactor N/A Rankine/
Brayton
Fast neutron reactors (Liquid metal cooled fast reactors)
40 Super PRISM 2280 N/A
41 BN-800 880 N/A
42 BN-600 600 N/A
43 FBR 500 N/A
44 BREST 300 Rankine
45 BN-350 350 N/A
46 STAR Secure transportable autonomous reactor N/A Brayton
47 PRISM Liquid metal cooled 150 N/A
48 SVBR Lead-bismuth fast reactor 75-100 Rankine
49 SSTAR Small sealed transportable autonomous
reactors
10-100 Brayton
50 LSPR Lead-bismuth cooled reactor 53 Rankine
51 ENHS Encapsulated nuclear heat source 50 Rankine
52 4S Super safe, small & simple, nuclear battery 10 & 50 Rankine
53 Rapid-L (under development) 0.2 Rankine
MSRs Molten salt reactors
54 AHTR Advanced high temperature reactor 1000 Brayton
55 FUJI MSR 100 Brayton


Generally, modern small nuclear reactors for power generation are expected to have greater simplicity of
design, economy of mass production and reduced siding costs. Many are also designed for a high level of
passive or inherent safety in the event of malfunction. Some are conceived for areas away from
transmission grids and with small loads, others are designed to operate in clusters in competition with
large units. Generation III+ designs are generally extensions of the generation III concept which include
advanced passive safety features. These designs can maintain the safe state without the use of any active
control components [16].


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752
Table 3. Capital cost estimates of generation III nuclear reactors

No Reactor Capacity (MWe) Cost (US$/kW)* Ref
1 EPR (Olkilmoto 3) 2x860 3341 [18]
2 EPR (Flamanville 3) 1600 3203 [18]
3 ABWR (Hitachi/Toschiba GE KK-6) 1315 2974 [19]
4 ABWR (Hitachi/Toschiba GE KK-7) 1315 2686 [19]
5 ESBWR (GE) 1560 1160-1250 [20]
6 APWR (Mitshubishi) 2x1700 1529 [21]
7 BWR 90+ (Westinghouse) 1650 1400 [22]
8 VVER-1500/V448 1500 1200 [23]
9 APR-1400 (South Korea) 1450 1400 [24]
10 ABWR (GE) 1326 1390 [25]
11 SWR-1000 1000-1290 1800 [26]
12 AP-1000 (Westinghouse Electric) 1100 1000 [25]
13 AP-1000 (Westinghouse) 1100 1200 [24]
14 VVER-91 (China) 2x1060 1245-1831 [27]
15 VVER-1000/V392 (Koodankulam) 2x1000 1500 [28]
16 VVER-640 645 1980 [29]
17 AP-600 (Westinghouse Electric) 600 1420 [25]
18 AP-600 600 166 [24]
19 IRIS 335 1000-1200 [24]
20 MSBWR 50 1950 [30]
21 IRIS-50 50 1950 [30]
22 KLT-40 (Severodrinsk) 2x40 4213 [31]
23 VBER-300 300 331 [32]
24 VK-300 (MED) 2x200 1140 [33]
25 ABV 2x38 3158 [32]
26 CAREM 300 1000 [34]
27 SMART 2x100 1615 [35]
28 SMART 1000 1800 [36]
29 SMART (South Korea) 65 6458 [37]
30 NP-300 (Technicatome) (MED) 300 442 [38]
31 NHR-200 (China) 200 552 [36]
32 CANDU-9 (Darlington) 4x881 3973 [39]
33 ACR-1000 1200 1000 [40]
34 AHWR 300 1176-1411 [41]
35 ACR-700 681 1000 [25]
36 GT-HTR (JAERI) 300 1300-1700 [16]
37 GT-MHR 288 972 [25]
38 GT-MHR (GA+Afrikantov) 285 1000 [16]
39 HTR-PM (Huaneng) 200 1500 [16]
40 PBMR (Escon) 165 108 [16]
41 Super PRISM 2280 1300 [42]
42 BN-800 800 1875 [43]
43 BN-600 560 10714 [44]
44 FBR 1250 4800 [45]
45 SVBR 75/100 661.5 [34]
46 4S (Toshiba+Criebi) 10 & 50 2500 [16]
47 AHTR 1000 1000 [16]
*Exchange rates used: 1 Japanese yen = 0.009593 US$, 1 Euro = 1.5531 US$, 1 tenge= 0.0082843 US$, 1 korean
Won= 0.0009763 US$, 1 crore= 10000000 Indian Rupee= 235127.93 US$.

4.1 Mitsubishi advanced pressurized water reactor
The Mitsubishi advanced pressurized water reactor (APWR) is a generation III nuclear reactor developed
by Mitsubishi Heavy Industries based on PWR technology. It features several design enhancements
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753
including a neutron reflector, improved efficiency and improved safety systems, including a combination
of passive and active systems. The core is surrounded by a steel neutron reflector which increases
reactivity. In addition, the APWR uses more advanced steam generators compared to the PWR, which
creates drier steam allowing for higher efficiency and more delicate turbines. This leads to an almost
10% efficiency increase compared to the PWR.
Several safety improvements are also notable. The safety systems have enhanced redundancy, utilizing 4
trains each capable of supplying 50% of the needed makeup water instead of 2 trains capable of 100%.
Also, more reliance is placed on the accumulators which have been redesigned and increased in size. The
improvements in this passive system have led to the elimination of the safety injection system, which is
an active system [17].

4.2 Advanced light water reactors
The advanced light water reactors (ALWR) incorporate all of the improved features of generation III but
have no real difference in terms of their operation between their generation II counterparts. The major
improvements are notably in the field of safety and improved economics [15, 46].

4.2.1 European pressurized reactor
The main design objectives of the European pressurized reactor (EPR) design are increased safety while
providing enhanced economic competitiveness through evolutionary improvements to previous PWR
designs scaled up to an electrical power output of 1600MWe. The reactor can use 5% enriched UO
2
or
uranium plutonium mixed oxide fuel. The EPR design has several active and passive protection measures
against accidents [18], such as, four independent emergency cooling systems, each capable of cooling
down the reactor after shutdown (i.e., 300% redundancy), leak tight container around the reactor, extra
container and cooling area if a molten core manages to escape the reactor. Ex-vessel cooling - corium
catcher and two-layer concrete wall with total thickness 2.6m, designed to withstand impact by airplanes
and internal over pressure.

4.2.2 Economic simplified boiling water reactor
The economic simplified boiling water reactor (ESBWR) is a passively safe generation III+ reactor
which builds on the success of the ABWR. Both are designs by General Electric Hitachi Nuclear Energy
(GEH), and are based on their BWR design. The ESBWR uses natural circulation with no recirculation
pumps or their associated piping. The passive safety systems in an ESBWR operate without using any
pumps at all, thereby further increasing design safety integrity and reliability, while simultaneously
reducing overall reactor cost. The technology also uses natural circulation for coolant recirculation within
reactor pressure vessel, therefore, there are no recirculation pumps and none of the associated piping,
power supplies, heat exchangers and instrumentation and controls.
ESBWR’s passive safety systems include a combination of systems that allow for the efficient transfer of
decay heat from the reactor to pools of water outside of containment. These systems utilize natural
circulation based on simple laws of physics to transfer the decay heat outside of containment while
maintaining water inventory inside the reactor keeping the nuclear fuel submerged in water and
adequately cooled.
The core is shorter than conventional BWR plants because of the smaller core flow, which is caused by
the natural circulation. There are 1132 bundles and the electric power can reach 1550MWe. Below the
vessel, there is a piping structure which allows for cooling of the core during a very severe accident.
These pipes divide the molten core and cool it with water flowing through the piping. The probability of
radioactivity release to the atmosphere is several orders of magnitude lower than conventional nuclear
power plants, and the building cost is 60-70% of other LWRs [47]. The energy production cost is lower
than other plants due to lower initial capital cost and lower operational and maintenance cost.

4.2.3 Advanced boiling water reactor
The advanced boiling water reactor (ABWR) is a generation III reactor based on the boiling water
reactor. The ABWR was designed by GEH and Toshiba. The ABWR generates electrical power by using
steam, which is boiled from water using heat generated by fission reactions within nuclear fuel, to power
a turbine connected to a generator. The standard ABWR plant design has a net output of about
1350MWe.
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The addition of reactor internal pumps mounted on the bottom of reactor pressure vessel achieves
improved performance while eliminating large recirculation pumps in containment and associated large-
diameter and complex piping interfaces with reactor pressure vessel. Only the reactor internal pumps
motor is located outside of reactor pressure vessel in the ABWR.
A fully digital reactor protection system with redundant digital backups as well as redundant manual
backups, ensures a high level of reliability and simplification for safety condition detection and response.
This system initiates rapid hydraulic insertion of control rods for shutdown when needed. Fully digital
reactor controls, with redundant digital backup and redundant manual backups, allow the control room to
easily and rapidly control plant operations and processes. Separate redundant safety and non-safety
related digital multiplexing buses allow for reliability and diversity of instrumentation and control. In
particular, the reactor is automated for startup (i.e., initiate the nuclear chain reaction and ascent to
power) and for standard shutdown using automatic systems only. Of course, human operators remain
essential to reactor control and supervision, but much of the busy-work of bringing the reactor to power
and descending from power can be automated at operator discretion.
The internal pumps reduce the required pumping power for the same flow to about half that required with
the jet pump system with external recirculation loops. Thus, in addition to the safety and cost
improvements due to eliminating the piping, the overall plant thermal efficiency is increased. Eliminating
the external recirculation piping also reduces occupational radiation exposure to personnel during
maintenance [48].

4.2.4 Advanced passive 1000 reactor
The advanced passive 1000 (AP1000) reactor is a two-loop PWR which produces a net of 1117MWe.
The safety systems apply passive protection, which is designed to yield such high degree of safety that
there is no need for the usual diesel generators, which provide the equipment with power in the case of a
loss of electrical supply. In the event of an accident they require little intervention, which reduces the
chance of human error and other failures. Safety enhancement is also achieved by using modern, reliable
devices. The probability of failures is further decreased by applying the concept of diversity (several
different types of systems are used and thus the effect of potential intrinsic failures can be avoided).
The design is less expensive to build partly due to the fact that it uses existing technology. The expense
is also reduced by rationalizing technology, which means decreasing not only the number of pipes, wires,
and valves necessary, but reducing a number of other components, and therefore reducing cost.
Standardization and type-related licensing will also help reduce the time and cost of construction. The
safety systems in the AP1000 are passive, relying on things like gravity and natural recirculation rather
than active systems such as pumps [49].

4.2.5 KLT-40C reactor
The KLT-40C reactor takes advantage of the experience gained through the operation of the KLT-40
reactors used to provide power for icebreaker propulsion. The KLT-40C utilises a reactor pressure vessel
and a loop nuclear steam plant configuration similar to a conventional PWR, incorporating forced reactor
coolant circulation at power. Four separate, helical coil, once-through steam generators and four canned
reactor coolant pumps are used. No boron is added to the reactor coolant during normal reactor operation.
Reactor pressure vessel and steam generators are adjacent to each other and at approximately the same
elevation, with concentric piping connections between reactor pressure vessel and the steam generators.
The capabilities of removing decay heat from both the primary and secondary system by natural
convection has been demonstrated experimentally. The steam generators and the reactor pressure vessel
are housed within cavities located in a metal-water shied tank. The air gap between the components and
the walls of the metal-water shield tank minimise heat loss to the shielding water during reactor
operation. Reactor pressure vessel cavity of the water-shield tank can be flooded under severe accident
conditions to prevent pressure vessel melt through. The KLT-40C is provided with a steel containment
structure. A secondary structure protects the containment from external events.
The major unique aspect of the KLT-40C nuclear power plant design is the incorporation of two KLT-
40C reactors into a comprehensive barge, referred to as the floating power unit. This consists of two
principal parts, living quarters and the process section. The living quarters provide all necessary living
accommodations for the operating staff. The process section houses the two KLT-40C reactors, the
control rooms, all other systems required for the normal operation of the power plants, and spent fuel and
radioactive waste storage facilities. Facilities to use the steam produced by the reactors for either
International Journal of Energy and Environment (IJEE), Volume 4, Issue 5, 2013, pp.743-776
ISSN 2076-2895 (Print), ISSN 2076-2909 (Online) ©2013 International Energy & Environment Foundation. All rights reserved.

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electricity production or process heat application are housed in separate on-shore facilities. Other on-
shore facilities include the switchyard, administration building, and the accident management centre [50].

4.2.6 CAREM reactor
The CAREM nuclear power plant has an integrated reactor. The entire high energy primary system-core,
steam generators, primary coolant and steam dome is contained inside a single reactor pressure vessel.
The flow rate in the reactor primary systems is maintained by natural circulation. The driving force
obtained by the differences in the density along the circuit are balanced by friction and form losses,
producing a flow rate in the core that allows for sufficient thermal margin to critical phenomena. The
coolant acts also as a moderator.
Self-pressurization of the primary system in the steam dome is the result of the liquid vapour
equilibrium, at which the core outlet bulk temperature corresponds to saturation temperature at primary
pressure. Heaters and sprinklers that are typical of conventional PWR’s are eliminated. Twelve identical
mini-helical vertical steam generators, of the once-through type are used to transfer heat from the
primary to the secondary circuit, producing dry steam at 47bar, with 30°C of superheating. The location
of the steam generator above the core induces natural circulation in the primary system.
The secondary system circulates upwards within the tubes, while the primary system does so in counter-
current flow (downward circulation). An external shell surrounding the outer coil layer, with an adequate
seal guarantees that the entire stream of the primary system flows through the steam generators. As
another safety feature, steam generators are designed to withstand the pressure from the primary system
up to the steam outlet and water inlet valves in case of loss of secondary pressure. The CAREM plant has
a standard steam cycle with a simple design. In accordance with the behavior of once-through boilers,
steam is superheated under all plant conditions and no super-heater is needed. Likewise, no blow-down is
needed in the steam generators, which reduces waste generation. A single turbine is used, and the exhaust
steam at low pressure is condensed in a water cooled surface condenser. The condensate is then pumped
and delivered to the full stream polishing system in order to maintain ultra-pure water conditions.
High purity water exiting the polishing system is sent to the low-pressure pre-heater using turbine
extraction as a heating medium. The warm water is delivered to the water accumulator in order to
perform degassing operations with additional heat using extraction steam. Water is then pumped to the
high-pressure pre-heaters (two in tandem using extraction steam) and sent to the steam generators as
feed-water, closing the circuit. The CAREM secondary circuit is not a safety-graded system, i.e., the
nuclear safety of the plant does not rely on the functioning of the steam circuit.

4.2.7 SMART reactor
The SMART reactor is an integral type power reactor with a rated thermal power of 330MW. It is
different from the loop-type reactors due to the arrangement of its primary components. All major
primary components, such as core, steam generators, pressurizer, and control element drive mechanisms,
and main coolant pumps, are installed in a single reactor pressure vessel. The integrated arrangement of
these components enables the elimination of large pipe connections between the components of the
primary reactor coolant systems, and thus fundamentally eliminates the possibility of large break
LOCAs. This integral arrangement, in turn, becomes a contributing factor to the safety enhancement of
the SMART. These innovative and advanced features are adopted in the SMART design to enhance its
safety, reliability, performance, and operability. Most of these technologies and design features
implemented in the SMART are those that have been well proven through the operation of commercial
power reactors, and new features will be proven through various tests.
Twelve identical steam generator cassettes are located on the annulus formed by reactor pressure vessel
and the core support barrel. Each steam generator cassette is of once through design with helically coiled
tubes wound around the inner shell. The primary reactor coolant flows downward in the shell side of the
steam generators tubes, while the secondary feed-water flows upward in the tube side. The secondary
feed-water exits the steam generator in a superheated steam condition. For performance and safety, each
steam generator cassette consists of six independent modules, and six modules from three adjacent steam
generators are then grouped into one nozzle. Three nozzles eventually compose one section. This concept
of steam generators grouping minimizes the asymmetric impact of a steam generator section isolation of
the reactor system.
An in-vessel self-pressurizing concept is adopted for the pressurizer of the SMART. The pressurizer is
located in the upper space of the reactor assembly and is filled with water and nitrogen gas. The concept
International Journal of Energy and Environment (IJEE), Volume 4, Issue 5, 2013, pp.743-776
ISSN 2076-2895 (Print), ISSN 2076-2909 (Online) ©2013 International Energy & Environment Foundation. All rights reserved.

756
of the self-pressurizing design eliminates the active mechanisms such as spray and heater. By keeping
the average primary coolant temperature constant with respect to power change, the large pressure
variation due to power change during normal operation can be reduced. To achieve self-pressurizing, a
pressurizer cooler for maintaining a low pressurizer temperature and a wet thermal insulator for reducing
heat transfer from the primary coolant are installed.
Main coolant pump is a canned motor pump that does not require any pump seal. This characteristic
eliminates a small break LOCA associated with a pump seal failure in the case of a station black out. The
SMART has four main coolant pumps installed vertically on reactor pressure vessel annular cover. Each
pump is an integral unit consisting of a canned asynchronous 3-phase motor and an axial flow single-
stage pump. A common shaft rotating on three radial and one axial thrust bearings connects the motor
and pump.
Besides the inherent safety characteristics of the SMART, further safety enhancement is accomplished
with highly reliable engineered safety systems. These systems are designed to function passively. The
shutdown of the reactor can be achieved by one of two independent systems. The primary shutdown
system is the control rods with Ag-In-Cd absorbing material. In the case of the failure of the primary
shutdown system, the emergency boron injection system is provided as an active backup. One of the two
trains is sufficient to bring the reactor to sub-critical condition [51].

4.2.8 MRX reactor
The MRX reactor is an integral style PWR with a thermal output of about 100MW, designed initially for
ship propulsion. Currently, other applications such as desalination and district heating are envisioned.
The MRX reactor size can be increased to about 300MWe without significant changes to the design
concept. An innovative feature of MRX is a compact steel containment vessel, which surrounds reactor
pressure vessel in relatively close proximity. The inter-space between reactor pressure vessel and the
containment vessel is water filled, with a nitrogen blanket in the top portion. As reactor pressure vessel
and other components that operate at elevated temperature in the inter-space between the containment
vessel and reactor pressure vessel are insulated to reduce heat loss, the insulation is protected by a
waterproof membrane.
Normal operating pressures in reactor pressure vessel and containment vessel are 120bar and 40bar
respectively. The two canned reactor coolant pump motors are each housed in horizontal canisters that
project from reactor pressure vessel above the core elevation, which serve to keep the motors isolated
from the containment vessel water. Hatches are provided in the containment vessel opposite the pumps to
facilitate inspection and maintenance. Although not shown in the submission, it is anticipated that a
shield building will be provided to protect the steel containment from external events.
The design of the MRX appears to offer several technical challenges. Among these is designing the
steam and feed-water lines to accommodate thermal expansion and seismic loads within the limited
space available, and the establishment and maintenance of the waterproof cladding over the insulation
that is applied to reactor pressure vessel, steam lines, feed water lines and other components that operate
at elevated temperatures. There appears to be a requirement for several pressure relief systems (for
example, for reactor pressure vessel, the containment vessel, and the canisters that house the reactor
coolant pump motors). In addition, vent lines are needed to equalise the pressures between the insulation-
filled cavities and the containment vessel. MRX incorporates pressurizer heaters conceptually similar to
those of conventional PWRs. Since the MRX employs a fuel cycle, a fuel design and fuel management
systems that are substantially the same as those of modern PWRs of conventional design, the general
characteristics regarding proliferation and safeguards application will be similar [50].

4.2.9 NHR-200 reactor
The NHR-200 reactor is a vessel type LWR with an integrated arrangement, natural circulation, self-
pressurized performance and dual vessel structure. The core is located at the bottom of reactor pressure
vessel. Primary heat exchangers are arranged on the periphery in the upper part of reactor pressure
vessel. The system pressure is maintained by inert gas and steam. A containment vessel fits tightly
around reactor pressure vessel, so that the core will not become uncovered under any postulated leakage
at the reactor coolant pressure boundary. There is a long riser on the core outlet to increase the natural
circulation capacity. The primary coolant absorbs the heat from the reactor core, then passes through the
riser and enters the primary heat exchangers, where the heat carried is then transferred to the intermediate