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Nuclear Engineering and Design 239 (2009) 1212–1219
Contents lists available at ScienceDirect
Nuclear Engineering and Design
journal homepage: www.elsevier.com/locate/nucengdes
Current status and technical description of Chinese 2 ×250 MWth HTR-PM
demonstration plant
Zuoyi Zhang, Zongxin Wu, Dazhong Wang, Yuanhui Xu, Yuliang Sun, Fu Li∗, Yujie Dong
Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084, China
article info
Article history:
Received 11 August 2008
Received in revised form 14 February 2009
Accepted 27 February 2009
abstract
After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great
efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with
much improved safety features. Following the successful construction and a most gratifying operation of
the 10 MWth high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy
Technology (INET) of Tsinghua University has developed and designed an HTR demonstrationplant, calle d
the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out
with industry partners from China and in collaboration of experts worldwide, closely follows the design
principles of the HTR-10.
Due to intensive engineering and R&D efforts since 2001, the basic design of the HTR-PM has been
finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR)
has been compiled.
The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one
comprising of a single zone 250 MWth pebble-bed modular reactor and a steam generator. The two NSSS
modules feed one steam turbine and generate an electric power of 210MW.
A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is
closely based on the technology of the HTR-10 fuel production line.
The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-
PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident
management procedures and will not require any need for offsite emergency measures.
According to the current schedule of the project the completion dateof the demonstration plant will be
around 2013. The reactor site has been evaluatedand approved; the procurement of long-lead components
has already been started.
After the successful operation of the demonstration plant, commercial HTR-PM plants are expected to
be built at the same site. These plants will comprise many NSSS modules and, correspondingly, a larger
turbine.
© 2009 Elsevier B.V. All rights reserved.
1. Introduction
After the nuclear accidents in Three Mile Island and Chernobyl
the world nuclear community started huge efforts to study and to
develop advanced nuclear power systems with enhanced safety
features. Advanced light water reactors (ALWR), which are lately
labeled Generation-III reactors (and even GEN-III+ reactors), have
been developed; e.g. the ABWR, the EPR, the System 80+, as well
as the AP1000 and ESBWR. LWR-concepts which are beyond the
ALWRs, such as IRIS by Westinghouse and PIUS by former ABB, were
also developed. In the field of high-temperature gas-cooled reactors
(HTGR), H. Reutler and G. Lohnert of German SIEMENS/INTERATOM
∗Corresponding author. Tel.: +86 10 62783838; fax: +86 10 62771150.
E-mail address: lifu@tsinghua.edu.cn (F. Li).
proposed the modular concept of a 200 MWth HTR-MODUL in the
early 1980s (Reutler and Lohnert, 1984). A very sophisticated, pecu-
liar design of the HTR-MODUL guarantees that the maximum fuel
temperature will never exceed the fuel’s design limit for all acci-
dents, such as e.g. “depressurized loss of coolant” or even “expulsion
of all control rods”, without needing any emergency cooling mea-
sures. According to statements made during an IAEA conference
in 1992 these kind of reactors were called “nuclear power plants
beyond the next generation”. Since 2000 these concepts were
labeled by the U.S. DOE as “Generation IV Nuclear Energy Systems”
(NERAC, 2002).
HTGRs use helium as coolant and graphite as moderator as
well as structural material. Its fuel elements contain thousands of
very small “coated particles” which are embedded in a graphite
matrix. At present, its core outlet helium temperature can reach
700–950 ◦C; even higher outlet temperatures are envisaged when
0029-5493/$ – see front matter © 2009 Elsevier B.V. All rights reserved.
doi:10.1016/j.nucengdes.2009.02.023
Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219 1213
the current research for better materials and improved fuel proves
to be successful. Therefore, HTGR plants are even now capable of
utilizing the high efficient and mature technologies of conventional
fossil-fired power plants. For example, HTGR plants can achieve
a thermal efficiency of 42% by even employing subcritical super-
heated steam turbines or reaching ∼45% when supercritical steam
turbines are installed. The efficiency could be improved even fur-
ther when adopting direct helium gas turbines with recuperators or
when choosing a combined cycle. In addition, the high-temperature
heat sources provided by HTGRs can be used in many industrialpro-
cesses to replace coal, oil or natural gas. Only some processes such
as heavy oil recovery, hydrogen production, coal gasification and
liquefaction will be mentioned here.
From the early 1960s, the United Kingdom, the United States
and Germany began to research and develop HTGRs. In 1962, the
U.K. and the European Community cooperated to build the first
HTGR (Dragon) in the world, which provided a thermal power of
20 MW and achieved criticality in 1964; it used fuel in the form of
graphite rod-bundles. Thereafter, Germany successively built two
pebble-bed nuclear plants, the 45 MWth test HTGR (AVR) and the
750 MWth HTGR power plant (THTR-300), while the U.Sconstructed
the 40 MWegraphite fuel rod-bundle core (the Peach-Bottom test
reactor) and the 330MWeFort. St. Vrain power plant utilizing pris-
matic graphite fuel. Japan started the construction of a 30MWth
high-temperature test reactor (HTTR) in 1991, which attained its
first criticality in 1998 (IAEA-TECDOC-1198, 2001).
Intensive R&D on modular HTGRs has been performed
in Germany and in the United States since the 1980s.
SIEMENS/INTERATOM designed a 200 MWth pebble-bed mod-
ular high-temperature gas-cooled reactor (HTR-MODUL), while
General Atomics adopted the principles of the German HTR-
MODUL and worked on a 350MWth prismatic fuel type modular
high-temperature gas-cooled reactor (MHTGR), which later on
was upgraded to a power output of 600MWth and connected to
helium gas turbine, the GT-MHR. Since the middle of the 1990s,
the company PBMR (pebble-bed modular reactor) of South Africa
is developing a 400 MWth pebble-bed modular reactor which also
adopts the helium gas turbine cycle. The electrical power is envis-
aged to be 165MW for an inlet-temperature of 500 ◦C and an outlet
temperature of 900 ◦C. In the framework of Generation-IV reactors,
the United States is planning to implement the Next Generation
Nuclear Plant (NGNP) project around 2020 by constructing an
HTGR demonstration plant for electricity and hydrogen production
(Idaho National Laboratory, INL/EXT-07-12967, 2007).
Fig. 1. Chinese 10 MW high-temperature gas-cooled test reactor (HTR-10).
1214 Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219
In China, R&D of HTGRs began in the middle of the 1970s.
From 1974 to 1985, the “Institute of Nuclear and New Energy
Technology” (INET) of the Tsinghua University carried out some
basic research on HTGR technology. After 1986, the R&D of HTGR
technology was highly intensified by co-operating with the inter-
national HTGR community, especially with German institutions.
During 1986–1990 eight key technical research topics, including 43
sub-tasks, were identified and systematical in-depth experimen-
tal studies were carried out. In 1992, China decided to construct
the 10 MWth high-temperature gas-cooled test reactor (HTR-10) as
shown in Fig. 1 (Wu et al., 2002) at the INET-site in Beijing.
This project is considered to be the first tangible step of HTGR
development in China. The main objectives of the HTR-10 were:
•to acquire the know-how to design, construct and operate HTGRs,
•to demonstrate the inherent safety features of the modular HTGR,
•to establish an irradiation and experimental facility for fuel ele-
ments and materials,
•to carry out R&D work for nuclear high-temperature process-heat
applications.
In June 1995, the first concrete for the HTR-10 reactor building
was poured. Finally, in December 2000, the HTR-10 reached its first
criticality, while in January 2003 the HTR-10 had been successfully
connected to the electric grid. For 72h the reactor worked at full
power. From April 2003 to September 2006 INET completed four
experiments to confirm and verify claimed crucial inherent safety
features of modular HTRs:
•loss of offsite power without any counter-measures,
•main helium blower shutdown without any counter-measures,
•loss of main heat sink without any counter-measures, and espe-
cially
•withdrawal of all rods without any counter-measures.
All these experiments were authorized, guided and supervised
by the National Nuclear Safety Authority (NNSA).
During this R&D-period of the HTR-10, five significant achieve-
ments were obtained:
(1) Manufacture of spherical coated particle fuel element: the know-
how of fabricating fuel elements for the HTR-10 was mastered.
The free Uranium fraction could steadily be decreased to a value
as low as 3 ×10−5.
(2) Corresponding technologies for pebble-bed HTGRs: the technology
of fuel element handling and spherical fuel element transporta-
tion by pulse pneumatic mechanism.
(3) Helium process technologies: such as helium sealing and purifi-
cation, the lubrication for rotating equipments in a helium
atmosphere, electrical insulation and rotor dynamics.
(4) Domestic manufacture of key equipments for HTGRs: this mainly
covers the reactor pressurevessel, the steam generator pressure
vessel, the hot gas duct, the once-through steam generator with
helical tubes, the helium blower, the fuel handling equipments
and reflector graphite components.
(5) Successful development of fully digital reactor protection systems.
The Chinese licensing authority, i.e. the National Nuclear Safety
Authority, issued all required licensing documents by carefully
and intensively reviewing the Preliminary Safety Analysis Report
(PSAR), the Final Safety Analysis Report (FSAR) and other relevant
supplementary documents. Thus, by licensing the HTR-10 the NNSA
acquired large experience and knowledge of HTGRs.
A second step of HTGR-application in China had been started
in 2001 (Zhang and Yu, 2002) when the high-temperature gas-
cooled reactor pebble-bed module (HTR-PM) project was launched.
The preliminary investment agreement was signed in December
2004 by “China Huaneng Group”, by “China Nuclear Engineering
and Construction Corporation” and by “Tsinghua Holding Cor-
poration”. In January 2006 the project named “Large Advanced
Pressurized Water Reactor and High-Temperature Gas-cooled Reac-
tor Nuclear Power Plants” became one of the 16 top priority
projects of the “Chinese Science and Technology Plan” for the
period 2006–2020. By February 2008 the implementation plan
and the budget for the HTR-PM project was approved by the
State Council of China. In November 2003, the “Chinergy Com-
pany” has been established and was designated to be the main
contractor of the HTR-PM nuclear island, while in January 2007
the “Huaneng Shandong Shidao Bay Nuclear Power Company”
was founded being the owner of the HTR-PM demonstration
plant.
2. Significance and technical progress of the HTR-PM
project
2.1. Major tasks of the HTR-PM technology in the Chinese market
The roles of the HTR-PM technology in the Chinese market are:
(1) Alternative to LWRs in nuclear power: the Chinese government
has announced to have 40 GWenuclear power plants in oper-
ation and additional 18 GWein construction by 2020. The
nuclear power capacity will be expanded to more than 100 GWe
between 2020 and 2040. The HTR-PM could be a supplement
to larger LWR plants.
(2) Alternative to oil and natural gas: China is currently the second
largest oil import country in the world. The HTR-PM could pro-
vide a high-temperature heat source for hydrogen production,
for heavy oil thermal recovery, for coal gasification and lique-
faction and for other industrial heat needs. The HTR would be a
major nuclear solution for these purposes.
(3) Next step of technology innovation after HTR-10: the HTR-10
had been the first step of HTGR development and of advanced
nuclear energy systems. The HTR-PM must – consequently – be
the next step, otherwise the gained vast expertise and the large
economic expenses during the last 20 years will be lost.
2.2. Main technical goals of the HTR-PM project
The HTR-PM should achieve the following technical goals:
(1) Demonstration of inherent safety features: the inherent safety
features of modular HTGR powerplants guarantees and requires
that under all conceivable accident scenarios the maximum
fuel element temperatures will never surpass its design limit
temperature without employing any dedicated and special
emergency systems (e.g. core cooling systems or special shut-
down systems, etc.). This ensures that accidents (e.g. similar to
LWRs core melting) are not possible so that not acceptable large
releases of radioactive fission products into the environment
will never occur.
(2) Demonstration of economic competitiveness: the first HTR-PM
demonstration power plant will be supported by the Chinese
government, so that the owner can always maintain the plant
operation and obtain investment recovery. However, this gov-
ernment supported demonstration plant has to prove that a
cost overrun during the construction period will be avoided and
that the predicted smooth operation and performance will be
kept. Hence, the demonstration plant must clearly demonstrate
that follow-on HTR-PM plants will be competitive to LWR plants
without any government support.
Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219 1215
(3) Confirmation of proven technologies: in order to minimize the
technical risks the successful experiences gained from the HTR-
10 and from other international HTGR plants will be fully
utilized in the HTR-PM project. The HTR-PM reactor design is
very similar to the HTR-10. The turbine plant design will use
the mature technology of super-heated steam turbines which
is widely used in other thermal power plants. Besides, the
manufacture of fuel elements will be based on the technol-
ogy verified and proven during the HTR-10 project. In addition,
the key systems and equipments of the plant will be rigorously
tested in large-scale experimental rigs in order to guarantee the
safety and reliability of all components. Furthermore, interna-
tional mature technologies and successful experiences will be
absorbed through international technical consultations.
(4) Standardization and modularization: the HTR-PM demonstra-
tion plant, consisting of two pebble-bed module reactors of
combined 2 ×250 MWth power, adopts the operation mode of
two modules connected to only one steam turbine/generator
set. This design allows to demonstrate the advantages and key
benefits of employing and implementing a design of standard-
ization and modularization. If the construction and operation
of the HTR-PM demonstration plant proves to be successful,
larger scale HTR-PM plants – using multiple-modules feeding
one steam turbine only – will become a reality.
2.3. Technical progress of the HTR-PM
The technical research for the HTR-PM began in 2001. The main
technical scheme of the nuclear island was finally fixed in 2006. The
key technology research and engineering verifications are carried
out according to elaborate plans. An HTGR engineering laboratory
and a large helium engineering testing loop are under construction
at INET. Here, the engineering verificationexperiments for the main
component prototypes will be performed on large test rigs offsite
the reactor.
The expected project construction period from pouring the first
tank of concrete to generating electricity for the grid is scheduled
to be 50 months. Although the workload of building, construction
and installation is relatively clear and straight forward, the project
schedule, nevertheless, leaves certain time margins allowing for
possible uncertainties. The current plan aims for feeding electricity
to the national power grid in 2013.
According to the requirements of the HTR-PM project, a fuel
production line will be built soon having a capacity of producing
300,000 spherical fuel elements per year.
Finally, the HTR-PM project will establish the technical foun-
dations to be able to realize Generation-IV nuclear energy system
goals in the next stage, such as:
(1) Largely enhanced safety features: a successful HTR-PM will have
already proven this technical target of Generation-IV nuclear
energy systems.
(2) Achieving outlet temperatures beyond 1000 ◦C [very high-
temperature gas-cooled reactor (VHTR)]: the reactor of current
design and using current fuel element technologies has already
the potential of realizing a gas outlet temperature of 950 ◦C.
A further improvement of the fuel element performance is
already foreseeable which will allow reaching this goal of
attaining an outlet-temperature of 1000 ◦C.
(3) Hydrogen production, use of helium turbine or supercritical steam
turbine: the current reactor design, verified by the HTR-PM,
can readily be applied for the helium turbine or super-critical
steam turbine or for the generation of large-scale production of
hydrogen by nuclear energy.
3. Technical description of HTR-PM
3.1. Overall technical description
The HTR-PM deploys pebble-bed modular high-temperature
gas-cooled reactors of 250 MW thermal power. Two reactor mod-
ules are coupled with two steam generators which are connected to
one steam turbine-generator of 210MW electric power. The reactor
and the steam generator are installed inside two separate pressure
vessels. The pressure vessels are assembled in a staggered, side-by-
side arrangement and are connected by a horizontal coaxial hot gas
duct. The primary pressure boundary consists of the reactor pres-
sure vessel (RPV), the steam generator pressure vessel (SGPV) and
the hot gas duct pressure vessel (HDPV), which all are housed in a
concrete shielding cavity as shown in Fig. 2.
The main helium blower is mounted on the upper part of the
steam generator pressure vessel. The core inlet helium temperature
was chosen to be 250 ◦C while the outlet helium temperature is
750 ◦C. The blower transfers the reactorheat to the steam generator,
where high-pressure super-heated steam is produced which drives
the steam turbine.
The ceramic structures surrounding the reactor core consist of
the inner graphite reflector and outer carbon brick layers. The reac-
tor core does not contain any fuel-free region or a graphite reflector
in the center. The control rod channels are located in the side
graphite reflector close to the core, while the returning cold helium
is guided through borings in the outer part of the side reflector. The
whole ceramic internals are installed inside a metallic core barrel,
which itself is supported by the RPV. The metallic core barrel and
the pressure vessel are protected against high temperatures from
the core by the cold helium borings of the side reflector, which act
like a shielding temperature screen.
The spherical fuel element with a diameter of 60 mm contains
a multitude of ceramic coated particles. The coated fuel particles
are uniformly embedded in a graphite matrix of 50 mm in diame-
ter; while an outer fuel-free zone of pure graphite surrounds the
spherical fuel zone for reasons of mechanical and chemical protec-
tion. Each spherical fuel element contains about 12,000 coated fuel
particles. A coated fuel particle with a diameter of nearly 1.0 mm is
composed of a UO2kernel of 0.5 mm diameter and three PyC layers
and one SiC layer (TRISO). The heavy metal contained in each spher-
ical fuel element is chosen to be 7.0 g. The design burn-up will be
90 GWd/tU, while the maximum fuel burn-up will not be in excess
of 100 GWd/tU. In order to reach a fairly uniform distribution of fis-
sile material throughout the whole core a “multi-pass” scheme of
fuel circulation had been adopted.
In summarizing, the HTR-PM has the following important tech-
nical design features:
(1) By using spherical fuel elements containing TRISO coated parti-
cles one can assure that all relevant radioactive fission products
will effectively be retained for at least 500 h when, during acci-
dents, a maximum temperature of 1620 ◦C is not exceeded.
(2) The pebble-bed core design allows the spherical fuel elements
to constantly pass though the core by gravity from up to down.
This fuelling scheme avoids loading the core with excess reac-
tivity. The elaborate reactor core design ensures that the fuel
element temperature will never exceed the safety limit of
1620 ◦C for any operating or accident condition.
(3) The operation mode adopts continuous fuel loading and dis-
charging: the fuel elements drop into the reactor core from
the central fuel loading tube and are discharge through a
fuel extraction pipe at the core bottom. Subsequently, the dis-
charged fuel elements pass the burn-up measurement facility
one by one. Depending on their state of burn-up they will
either be discharged and transported into the spent fuel stor-
1216 Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219
Fig. 2. Cross-section of the HTR reactor building.
age tank when having reached their design burn-up, or they
will be re-inserted into the reactor to pass the core once again.
The power distribution of the core depends on the number of
passes one chooses. Obviously, the higher the number of passes
is chosen the flatter will be the power distribution. This is favor-
able when regarding fuel element temperaturesduring accident
conditions. On the other hand, a high number of fuel passes
complicates the fuel handling devices as well as the complexity
of the burn-up measurement facility.
(4) Two independent shutdown systems are installed: a controlrod
system and a small absorber sphere (SAS) system, both placed
in holes of the graphitic side reflector. For shutdown purposes
the rods and the small absorbers are released and drop into the
reflector borings by gravity. This will improve the reliability of
the shutdown systems.
(5) The active core zone is encased by a bulky layer of graphite and
carbon bricks without metallic components. This ensures that
the core internals can withstand and endure very high temper-
atures.
(6) The reactor core and the steam generator heat transfer bun-
dles are installed in two different pressure vessels, which are
connected by the hot gas duct pressure vessel. The primary
pressure boundary comprises all three vessels. These vessels
are all protected by the cold helium gas (250◦C). This ensures
that moderate vessel temperatures are reached during reactor
operation and in accident scenarios.
(7) All three primary loop pressure vessels (RPV, SGPV and the
HDPV) are located in a concrete cavity, which protects the pri-
mary loop from external loads.
3.2. Main reactor plant equipments
The main systems and equipments of the reactor plant include
the reactor internals, the control rod system and the small absorber
sphere system, the reactor primary pressure vessels, the main
helium blower and the steam generator.
The reactor internals consist of graphitic, carbonic and metallic
components. The graphitic internals act primarily as the neutron
reflector; in addition they provide the means to be able to arrange
the helium flow channels and the absorber borings. The main func-
tion of the metallic internals is to support the graphite and carbon
internals along with the ceramic structure of the pebble-bed core,
and to pass various loads and forces to the reactor pressure vessel.
The control rod system and the small absorber sphere system are
two independent control systems of reactivity. These two indepen-
dent systems fulfill the requirements of diversity and redundancy.
There are 8 control rods and 22 small absorber sphere units, both
are located in the reflector region.
The three primary pressure vessels are composed of SA533-B
steel as the plate material and (or) the 508-3 steel as the forg-
ing material. These materials meet the technical requirements of
ASME-III-1-NB.
Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219 1217
Fig. 3. Cross section of steam generator with 19 assemblies.
The main helium blower, designed as a vertical structure, is
installed on the top of the steam generator inside the steam gener-
ator pressure vessel. The electric motor is mounted on an insertable
assembly; the motor is driven by the converter outside of the pres-
sure vessel. A magnetic bearing system is envisaged.
The steam generator consists of 19 separate helical tube assem-
blies; each assembly has 5 layers and includes 35 helical tubes, as
shown in Figs. 2 and 3. To ensure two-phase flow stability, throt-
tling apertures are installed at the entrance of all helical tubes.
The assembly type design of the steam generator uses the experi-
ences from the steam generator employed in the HTR-10. In-service
inspection is possible. For full verification of the steam generator
assembly full scale testing will be performed.
Due to the favorable temperature features of HTGRs, a super-
heated high-pressure steam turbine is adopted for the HTR-PM;
these components exhibit high reliability and economical viability.
Above all, there are mature experiences in the design, fabrication,
serial manufacturing and operation of these systems and compo-
nents in China.
3.3. Main technical parameters
The main technical parameters of the HTR-PM are presented in
Table 1.
3.4. Key technical decisions
Before deciding on the design of the HTR-PM some fundamental
decisions were made:
(1) Steam turbine cycle or helium turbine cycle:from2001to2003,
INET cooperated with the East China Electric Power Design
Institute (ECEPDI) to carry out a pre-conceptual research for
HTGR power plants. Three technical solutions were compared
in detail; these included:
(a) the conventional indirect steam turbine cycle,
(b) the direct helium turbine arrangement, and
(c) the indirect helium turbine arrangement.
Table 1
Main design parameters of the HTR-PM.
Parameter Unit Value
Rated electrical power MWe210
Reactor total thermal power MWth 2×250
Designed life time a 40
Average core power density MW/m33.22
Electrical efficiency % 42
Primary helium pressure MPa 7
Helium temperature at reactor
inlet/outlet
◦C 250/750
Fuel type TRISO (UO2)
Heavy metal loading per fuel
element
g7
Enrichment of fresh fuel
element
% 8.9
Active core diameter m 3
Equivalent active core height m 11
Number of fuel elements in one
reactor core
420,000
Average burn-up GWd/tU 90
Type of steam generator Once through helical coil
Main steam pressure MPa 13.24
Main steam temperature ◦C566
Main feed-water temperature ◦C 205
Main steam flow rate at the
inlet of turbine
t/h 673
Type of steam turbine Super high-pressure
condensing bleeder turbine
1218 Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219
It was found that the direct helium turbine technology had
certain technical uncertainties. Thus, the conditions for con-
structing an industrial demonstration plant of this type were
not ready yet. The following technical problems needed to be
solved: the inspection technology of helium turbo-compressor
blades under aggravating conditions of radioactivity deposition,
the design and verification technology of RPV materials, the
magnetic bearing technology, high-speed rotor dynamics and
control technology, high efficiency recuperator-technology, to
name only a few.
As for the R&D issue of helium turbine technology, INET
is now in charge of the national research project named
“10MW High-Temperature Gas-cooled Reactor Helium Tur-
bine System” (HTR-10GT). The goal of this project is to
build a helium turbine electricity-generating system to study
the key technical difficulties related to the HTGR helium
turbine technology. The commercial-scale HTGR helium tur-
bine cycle could only be realized in the future when a
debugged, tested and verified HTR-PM reactor had been com-
bined with mature and verified components of helium turbine
technology.
In 2003 it was decided to use a steam turbine cycle for the
HTR-PM project after the three above mentioned cycles had
been intensively studied and scrutinized. The plant will have
a design thermal efficiency of 42%.
(2) Two-module reactors coupled with one steam turbine: two differ-
ent designs have been studied –
(a) The first one was a 458 MWth reactor with a two-zone annu-
lar core (Zhang et al., 2006). This kind of design adopts
a fuel-free graphite reflector placed in the center of the
pebble-bed core in order to increase the reactor thermal
power as much as possible. However, our evaluations indi-
cated that – for the time being – there are grave technical
uncertainties in this design.
(b) The second design consists of a 2 ×250 MWth reactor plant;
each reactor has a one-zone cylindrical core. The Chinese
HTR-10 is a one-zone pebble-bed reactor in a side-by-side
arrangement of rector and steam generator. Hence, the
design of a 2 ×250 MWth plant of two one-zone reactors
is regarded as a mere up-scaling of the HTR-10 as proto-
type. Through the practices and experiences obtained by the
design, construction and operation of the HTR-10, the tech-
nical uncertainties of this second design will be reduced
decisively. Besides, the HTR-MODUL plant, designed by
SIEMENS/INTERATOM and licensed by the German licensing
authorities, also adopted the two-module reactors design
and will be regarded as a reference.
(3) Costs: after having carried out the economic comparison of the
above first two designs, it wasclearly found that the specific cost
differences are small (Zhang and Sun, 2007). The main reasons
are as follows –
(a) In order to reduce the helium flow resistance, the primary
pressure of the 458 MWth reactor had to be increased to
9.0 MPa. Also the diameter of the pressure vessel had to
be enlarged. By contrast, the 250 MWth reactor needs only
a pressure of 7.0 MPa while having a smaller diameter of
the pressure vessel. Therefore, the total weight of the pri-
mary pressure boundary components of the 2 ×250 MWth
reactors is only 14% higher than the weight of the primary
pressure boundary of a 458 MWth reactor.
(b) Three trains of equipments for the fuel handling systems are
required for the 458 MWth reactor, while the 2×250 MWth
reactors need only two.
(c) Since one has to take into account a necessary replacement
of the central graphite reflector during the lifetime of the
458 MWth plant, the reactor building for this design is higher
Table 2
Comparison of two HTR-PM designs.
1×458 MW design 2×250 MW design
RPV weight 1 2 ×0.57
Graphite weight 1 2 ×0.60
Metallic reactor internals weight 1 2 ×0.86
Main blower power 1 2 ×0.57
Number of control rods 24 2 ×8
Number of small absorber sphere systems 8 2 ×22
Number of fuel handling system 3 2
Volume of reactor plant building 1 0.96
Number of reactor protection systems 1 2
Number of main control room 1 1
Helium purification systems 2 ×100% 2 ×10 0%
Fresh fuel and spent fuel systems 1×100% 1 ×10 0%
Emergency electrical systems 2 ×100% 2 ×10 0%
and larger. The detailed comparison of the two designs is
depicted in Table 2.
(4) Other aspects: the “module” concept has the following meaning
–
(a) aeveral identical modules are arranged to form a large plant,
which is of benefit to standardization as well as reduction
of manufacturing costs;
(b) a relatively small nuclear power-output of a module reac-
tor is indispensable when wanting to realize inherent safety
features;
(c) therefore, a modular nuclear power plant must consist of
several or even many modules. However, the total plant
capacity, but not the single module power, is pivotal and
decisive for the economics of a power plant. Auxiliary sys-
tems, infrastructure and other indirect costs can be shared
by all the modules.
The HTR-PM adopts a layout mode of two-module reactors
coupled to one steam turbine in order to verify the feasibil-
ity and rationality of coupling a multitude of modules to one
steam turbine. In addition, this kind of arrangement enables
to test the sharing of the auxiliary systems. Furthermore, a
2×250 MW thermal power output coincides with the availabil-
ity of 200 MWesteam turbine products in the Chinese market.
4. Safety-performance and economics of the HTR-PM
4.1. Safety-performance of the HTR-PM
The HTR-PM will realize the following safety features:
(1) the radioactive inventory in the primary helium coolant is very
small when the reactors are working at normal operation con-
ditions. Even if this limited amount of radioactivity would be
released into the environment following an accident, there is
no need to take any emergency measures;
(2) for any reactivity accident or for a loss of coolant accident the
rise of the fuel element temperature will not cause a signifi-
cant additional release of radioactive substances from the fuel
elements;
(3) the consequences of water or air ingress accidents depend on
the quantity of such ingresses. The ingress processes and the
associated chemical reactions are slow, and can readily be ter-
minated within several dozens of hours (or even some days) by
taking very simple actions.
The nuclear safety goal of the HTR-PM can be summarized fol-
lows:
The consequences of all conceivable accidents will not result
in significant offsite radioactive impacts. The plant meets already
the safety target of Generation-IV nuclear energy systems which
Z. Zhang et al. / Nuclear Engineering and Design 239 (2009) 1212–1219 1219
stipulates: “eliminate the need for offsite emergency measures”.
The same viewpoint is put down by IAEA in its report No. NS-R-
1 “Safety of Nuclear Power Plants: Design”. Here it is expressively
stated that “An essential objective is that the need for externalinter-
vention measures may be limited or even eliminated in technical
terms, although such measures may still be required by national
authorities.”
4.2. The economics of the HTR-PM
According to our investigations and regarding specific costs
(Zhang and Sun, 2007), there is no significant difference between
an HTR-PM plant and a PWR plant when the costs of infrastruc-
ture, R&D, project management, etc. are effectively shared in a
commercial-scale, multiple-module HTR-PM plant. Compared with
PWRs, inherently safe HTR-PM plants exhibit smaller power den-
sity, in total heavier PRVs and core internals, and higher specific
cost. The other components of a nuclear power plant, however,
depend upon the power to be generated, and no significant dif-
ference exists between PWRs and an HTR-PM plant. The reactor
pressure vessel and the costs of reactor internals of a PWR accounts
for only ∼2% of the total plant costs (including financial cost, from
the practical data in Chinese PWR project, Zhang and Sun, 2007),
so the cost increase from RPVs and reactor internals in HTR-PM has
a limited impact. This limited impact will be compensated by sim-
plification of the reactor auxiliary systems, the I&C and electrical
systems, as well as by the benefit of mass productionfor the conven-
tional island equipments, RPVs and reactor internals. In addition,
it is expected that the costs of an HTR-PM plant will be further
decreased through reducing the workload of design and engineer-
ing management, shortening construction schedules and lessening
financial costs by making use of modularization.
In summing up it is expected that modular HTGR power plants
will show to be economically competitive with PWRs due to the
following reasons:
(1) simple systems;
(2) high operation temperature and the use of a high-pressure
super-heated steam turbine-generator; this is similar to normal
fossil power plants. Hence, a much higher thermal efficiency can
be realized;
(3) multiple-module reactors coupled to one steam turbine-
generator, sharing common auxiliary systems, and further
reducing the costs through modularization and standardization
for manufacture and construction;
(4) the operation mode of on-line continuous fueling will improve
the availability of the power plant;
(5) the design burn-up of the fuel is expected to reach at least
100 GWd/t or even more; this will reduce the fuel cycle costs.
From our current knowledge and for Chinese market conditions
we estimate the necessary budget excluding R&D and infrastruc-
ture costs for the first HTR-PM demonstration plant to be about
2000 USD/kWe.
Of course, all these claims, drawn from our year-long analysis,
must clearly be verified in detail. By successfully operating the HTR-
PM in the very near future we are confident to reach these our
claims.
5. Conclusions
On the basis of the HTR-10, the ongoing Chinese HTR-PM project
is considered to be a decisive new step for the development of
Chinese HTGR technology. Its main target is to finish building a
pebble-bed HTR-PM demonstration plant of 210 MWearound 2013.
Through the mutual efforts of all relevant scientific research orga-
nizations and industrial enterprises, and having the strong support
of the Chinese government, the HTR-PM project will certainly play
an important role in the world-wide development of Generation-IV
nuclear energy technologies.
References
Idaho National Laboratory, 2007. Next Generation Nuclear Plant, Pre-Conceptual
Design Report, INL/EXT-07-12967.
International Atomic Energy Agency, 2001. Current Status and Future Development
of Modular High Temperature Gas Cooled Reactor Technology, IAEA-TECDOC-
119 8 .
The United States Department of Energy’s Nuclear Energy Research Advisory Com-
mittee (NERAC) and the Generation IV International Forum (GIF), 2002. A
Technology Roadmap for Generation IV Nuclear Energy Systems.
Reutler, H., Lohnert, G.H., 1984. Advantages of going modular in HTRs. Nuclear Engi-
neering and Design 78, 129–136.
Wu, Z., Lin, D., Zhong, D., 2002. The design features of the HTR-10.Nuclear Engineer-
ing and Design 218, 25–32.
Zhang, Z., Yu, S., 2002. Future HTGR developments in China after the criticality of
the HTR-10. Nuclear Engineering and Design 218, 249–257.
Zhang, Z., Wu, Z., Sun, Y., Li, F., 2006. Design aspects of the Chinese modular high-
temperature gas-cooled reactor HTR-PM. Nuclear Engineering and Design 236,
485–490.
Zhang, Z., Sun, Y., 2007. Economic potential of modular reactor nuclear power
plants based on Chinese HTR-PM project. Nuclear Engineering and Design 237,
2265–2274.