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IAEA-TECDOC-479
STATUS OF
ADVANCED TECHNOLOGY AND DESIGN
FOR WATER COOLED REACTORS:
LIGHT WATER REACTORS
A TECHNICAL DOCUMENT ISSUED BY THE
INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1988
STATUS OF ADVANCED TECHNOLOGY
AND DESIGN FOR WATER COOLED REACTORS:
IAEA, VIENNA, 1988
Printed by the IAEA in Austria
Water reactors represent a high level of performance and safety. They
are mature technology and they will undoubtedly continue to be the main
stream of nuclear power. There are substantial technological development
programmes in Member States for further improving the technology and for the
development new concepts in water reactors. Therefore the establishment of
an international forum for the exchange of information and stimulation of
international co-operation in this field has emerged.
In 1987 the IAEA established the International Working Group on
Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the
framework of IWGATWR the IAEA Technical Report on Status of Advanced
Technology and Design for Water Cooled Rectors, Part I: Light Water Reactors
and Part II: Heavy Water Reactors has been undertaken to document the major
current activities and different trends of technological improvements and
developments for future water reactors. Part I of the report dealing with
LWRs has now been prepared and is based mainly on submissions from Member
States, and the Agency would like to thank all those individuals and
institutions who have contributed to it. In particular the Agency would
like to express its gratitude to the consultants, (see attached list) who
continuously reviewed the progress of the report and thus contributed
substantially to its successful completion. Thanks are also given to the
secretaries from the Agency's Division of Nuclear Power, who devoted to the
typing of the report.
It is hoped that this part of the report, containing the status of
advanced light water reactor design and technology of the year 1987 and
early 1988 will be useful for disseminating information to Agency Member
States and for stimulating international cooperation in this subject area.
Part II of the report dealing with HWRs is in preparation with release
expected during 1989.
In preparing this material for the press, staff of the International Atomic Energy Agency
have mounted and paginated the original manuscripts and given some attention to presentation.
The views expressed do not necessarily reflect those of the governments of the Member States
or organizations under whose auspices the manuscripts were produced.
The use in this book of particular designations of countries or territories does not imply any
judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of
their authorities and institutions or of the delimitation of their boundaries.
The mention of specific companies or of their products or brand names does not imply any
endorsement or recommendation on the part of the IAEA.
Please be aware that all the Missing Pages
in this document were originally blank pages
TRENDS IN ADVANCED LWR DESIGN AND TECHNOLOGY .........................
Incentives for the development of advanced LWRs ......................................
Design objectives for advanced LWRs ......................................................
Outline of the development of advanced LWRs ..........................................
Low temperature nuclear heat reactor .......................................................
IAEA programme on advanced light water reactors .....................................
PROGRAMME FOR ALWR DEVELOPMENT .................................................
Finland ............................................................................................
France .............................................................................................
Federal Republic of Germany ................................................................
Japan ...............................................................................................
Sweden ............................................................................................
United Kingdom .................................................................................
Union of Soviet Socialist Republics .........................................................
United States of America ......................................................................
CMEA member countries .....................................................................
Developing countries ...........................................................................
LARGE SIZE ALWR DESIGNS (above 600 MWe)
The N4 model (France) ........................................................................ 26
Convertible spectral shift reactor (RCVS) (France) ...................................... 32
The Convoy plants (PWR 1300) (Federal Republic of Germany) ..................... 36
The Siemens 1000 MWe three loop PWR (Federal Republic of Germany) ......... 39
High converter reactor (HCR) (Federal Republic of Germany) ........................ 44
Advanced BWR 90 (Sweden) ................................................................. 48
The VVER-1000 and VVER-1800 design (USSR) ....................................... 57
Combustion Engineering System 80 Plus (USA) ......................................... 70
Mitsubishi-Westinghouse (M-W)
APWR design (Japan-USA) ........................ 79
Hitachi-Toshiba-GE ABWR (Japan-USA) ................................................ 86
The Sizewell-B reactor (UK-USA) .......................................................... 99
Standard nuclear power design (PUN) (Italy) ............................................. 103
Experience with light water breeder reactor development and operation (USA) ... 106
MEDIUM SIZE ALWR DESIGNS (-600 MWe)
The development of MALWRs (medium ALWRs) .......................................
Finnish alternative (Finland) ..................................................................
B&W B-600 PWR (USA) .....................................................................
CE Minimum Attention Plant (USA) ........................................................
GE SBWR (simplified/safe BWR) (USA) ..................................................
Westinghouse AP-600 APWR (USA) .......................................................
Westinghouse NUPACK plant (USA) .......................................................
PIUS TYPE REACTORS ............................................................................. 139
5.1. SECURE-P reactor conceptual design (Sweden) .......................................... 139
5.2. ISER reactor (Japan) ........................................................................... 152
5.3. PIUS BWR and PECOS-BWR (USA) ...................................................... 156
ANNEX I.
SUMMARY TABLES FOR ALWRs ..................................................... 167
ANNEX II. SUMMARY TABLES FOR LOW TEMPERATURE HEAT REACTORS ...... 179
REFERENCES ................................................................................................
LIST OF CONSULTANTS ....................................................................................189
1. TRENDS IN ADVANCED LWR DESIGN AND TECHNOLOGY
INCENTIVES FOR THE DEVELOPMENT OF ADVANCED LWRs
Status of Current LWRs fl]
By the end of 1987, nuclear power generated about 16% of the
electricity worldwide. There were 417 nuclear plants in operation, 337
(~74%) were light water cooled reactors; 54% of nuclear power is produced
by PWRs. Nations such as France, the Federal Republic of Germany, Japan,
Belgium and Sweden are already heavily reliant on nuclear power. A very
broad nuclear power experience is available in the United States of America
which has over 100 operating nuclear power units. Nuclear power has
established its position as a viable alternative energy source in the
world. The current Light Water Reactor technology is a mature and proven
one, which had tremendous progress and consolidation in the last decades.
The development of the nuclear energy has reached a very high standard in
reliability and availability, and a very high level in performance and
safety. Since 1984, about 40% of the units are consistently reporting an
availability of more than 80%. The high load follow capability of LWR
plants is fully compatible with conventional power plants. The nuclear
generation cost is compatible with coal and will be cheaper than coal in
some regions. A stable construction period of 5-6 years has been
demonstrated and even a considerable reduction of that to about 4 years has
been realized. The high quality of operation and maintainance has been
reached in compliance with the stringent safety requirements, incorporating
the feedback from operation experience and the lessons learned from the
incidents and accidents.
In some countries the demand for electricity and nuclear power is lower
than it was originally expected. Also other considerations, such as high
construction costs and long construction periods of some nuclear power
plants, and more recently the concern about nuclear safety for severe
accidents and radioactive release in the existing reactors, have resulted in
a slowdown and re-examination of the nuclear option in some countries. But
it can still be expected that the nuclear share of the world's electricity
will be increased to about 20% by the year 2000. The programmes for nuclear
power as an increasingly important energy source are continued in
industrialized and some developing countries, such as France, Japan, the
USSR and the UK, India, Korea and China. There are substantial research and
development programmes in some Member States for further improving the
technologies and for the development of new design concepts in the Light
Water Reactor. It is foreseeable and undoubted that the Water Cooled Reactor
will be the main stream of nuclear power among all the lines of nuclear
reactor types in the next decades in the world.
Incentives for Development of Advanced LWRs [2-4]
As mentioned above, LWRs offer a broadly developed and mature
technology basis, and have a potential for further improvement. Various
advanced concepts, designs and technologies emphasize plant reliability,
availability and safety as well as economy. There are different directions
under consideration for LWR technology improvements and developments.
Some countries are aiming at better fuel utilization in current water
reactors. Because the plutonium stock in the late 1990s from the
reprocessing will considerably increase, and since plutonium can be best used
in fast reactors, one possible long-term strategy of using this plutonium
would be to build fast breeder reactors. But the large scale introduction of
the fast breeder reactor is not expected to be realized before 2010. The
role of LWRs as the main nuclear energy source for electricity will therefore
be increased and prolonged to meet the needs, including replacement of aged
decommissioned plants. A near-term target is increasing fuel discharge
burnup and using plutonium in existing LWRs. A future way might be the
introduction of tight lattice core with high conversion ratio, in which only
minor modifications over the existing LWRs are required and mainly are in the
reactor core and related components.
Some countries have adopted an evolutionary approach to developing LWR
plants with enhanced passive safety features, simplifications and
préfabrications for the 1990s. The improvements are being based on the
feedback from long operation experience with LWRs and results from R&D
programmes related to those systems. For these approaches, it is not
necessary to build demonstration plants and to conduct long-term development
programmes.
There are also initiatives for the development of new concepts, the
so-called "Inherently Safe" concepts, which can be called a developmental
approach. The PIUS (process inherent ultimate safety) and the ISER
(intrinsically safe and economical reactor) designs, in which no core melt
sequence has been identified, are the typical examples for these concepts.
DESIGN OBJECTIVES OF ADVANCED LWRs
The design objectives of Advanced LWRs emphasize plant safety,
reliability and availability as well as cost reduction in construction,
operation and maintenance.
Safety [5-7]
The safety of operating plants has been periodically improved by
backfitting from operating experience feedback and incorporation of advances
in technology development. Nuclear power plants operating today have
incorporated to a very large extent the lessons learnt from the incidents and
accidents. The reactor systems, components and engineered safety systems
have become very reliable.
For new plant designs, there are also various options to extend
desirable approaches for plant safety and for further reduction of residual
risk for nuclear power plant accidents, mainly for core melt accidents and
for radioactive release to the environment. One option includes passive
safety systems which are conceived to be very reliable and which depend on
gravity, thermohydraulics and reactor physics laws, and not requiring the
intervention of operators or the use of externally actuated electrical or
mechanical devices. Another option includes measures for increasing safety
design margins which include lower power density in the core, and larger
water inventory in the loops. Then the primary system and power plant would
have a longer response time and be less sensitive to plant abnormal initiative
events, transients and perturbation. New nuclear power plants are expected to
have new man-machine interface systems based on computerized instrumentation.
The public risk from radioactive release to the environment might be
further reduced or virtually eliminated for current and future reactors. The
controlled containment venting system is being applied in several countries.
Measures maintaining containment integrity in case of serious overpressure as
a consequence of a core melt accident, while confining the great majority of
fission products and retaining molten core material, are under consideration.
Plant Cost [8]
The competitiveness of nuclear power with alternative power generation
is an important factor in nuclear power development. The nuclear electricity
costs in different countries vary widely. In some countries, nuclear
generated electricity costs much less than the electricity from conventional
plants. In general, nuclear power has a clear advantage over coal for
baseload electricity generation in many countries. While in some countries
there are cheaper coals available near the load centers, and/or extensive
infrastructures requiring additional investments, then nuclear power could be
less favourable.
Further cost reduction of nuclear generated electricity from new power
plants to be built in the near future is expected. In order to achieve cost
reductions, plant construction schedules could be shorter, the licensing and
regulation made more predictable, construction management improved, and
construction techniques upgraded, i.e. automated welding and testing, shop
préfabrication of integrated package of equipments, entire subsystems, etc.
In some countries, the construction period for ALWR is expected for 4~5
In addition an improvement in plant economics can be achieved by better
fuel utilization which could significantly reduce the amount of uranium
requirements and separative work units. In some ALWR designs, with once
through fuel cycle, fuel burnup extension, spectral shift control with
mechanical water displacer rods, or fuel cycle length extension could reduce
the fuel cycle cost by 20%, save U-238 resources of about 20% and enrichment
work of 30% in comparison with existing LWRs. For spent fuel reprocessing
strategy and plutonium utilization, a conversion ratio of around 0.9 is
achievable in a convertible spectral shift reactor, or one with a tight
lattice core, and with the use of the thorium-U-233 cycle, breeding can be
achieved. Thus, the cost in the fuel cycle could be reduced substantially.
Other measures could further reduce the plant cost in investment, operation
and maintenance, including:
extention of plant design lifetime up to 60 years,
possible replacement of components which may shorten the operating
period, such as pumps, motors, actuators, I&C systems even to RPV,
shorten planned outages and prevent unplanned outages by the use of
automatic remote controlled inspection equipment with incorporated
intelligent electronic systems, 20-25 days refueling outage is
achievable,
increase plant thermal efficiency,
design simplification in systems and operation.
Plant Performance [9-11]
The operating plant performance has already reached today very high
figures. The new designs for ALWRs to be constructed in the 90's have more
concern with the plant performance in availability, reliability, operability
and maintainability. For this purpose, the design improvements not only
focus on the nuclear steam supply system, but also on the entire power plant
with its multi-face interactions. The plant availability for ALWRs is
aiming at high than 90% in some countries, and the planned outage time at
about 20~30 days/yr. on average.
The new design of PWR steam generator configuration, the new material
of steam generator tube and optimum water chemistry, show that the steam
generator problems which have led to great concern all over the world are
being handled properly. Some designs use a canned motor pump as the primary
coolant pump instead of a shaft seal pump. The canned motor pump has a
demonstrated track record of high reliability, and inherently reduces the
potential for small LOCA. These measures are examples to improve the
reliability of ALWRs.
Some designs adopt a new control system in order to increase load
follow capability and plant operability. The development of new maintanance
devices and improved designs for easier access to equipment inside the
containment increase the maintainability. Some designs of ALWRs using
large-piece forgings for the reactor pressure vessel and bend pipings in
place of elbows etc. drastically reduce the welds in the vital components.
Therefore, not only the inspection time can be reduced, but also the
equipment reliability will be improved. For the long-term development
inspection-free instruments, equipments and even inspection-free operation
may be achievable, using corrosion and abrasion resistant new materials.
The occupational radiation exposure for operation personnel has been
continuously reduced for operating plants and has reached a very
satisfactory low level by using remote controlled and automated inspection
equipment and the respective counter-measures in the plant design. The more
stringent target set in some countries is less than 0.5 manSv/yr.
OUTLINE OF THE DEVELOPMENT OF ADVANCED LWRs (ALWR)
The trends in Advanced LWR design and technology in the next decades
seem to direct towards fuel utilization, evolutionary improvement of plants,
as well as innovative designs and concepts.
Improvement of Fuel Utilization
Spectral shift high conversion reactors are described in section 3.2
and 3.5. The concepts relate to a tight lattice reactor core, and are in
the feasibility study and in the R&D phase. It is a new way to provide the
flexibility of fissile material usage, not only for enriched uranium fuel
with a reduction of once-through fuel cycle cost compared to the current
LWR, but also for plutonium produced from reprocessing or mixed uranium and
plutonium oxide (MOX) fuel. The conversion ratio could reach ~ 0.9. The
reactor core and related reactor internals will be of a rather innovative
design, which might make a test programme including verification facilities
necessary. The other parts will be based on existing LWR plants and will
only need minor modification.
The Mitsubishi-Westinghouse (M-W) APWR, which is planned to complete
the construction in the late 90s, uses a spectral shift control system with
water displacer rods (section 3.9). Along with other measures, e.g. the low
power density core, Zircaloy grids, and the radial reflector etc., a saving
of 23% in U-238 resources and 30% in enrichment work can be achieved.
Breeding in LWRs is possible and has been demonstrated through an
extensive program, utilizing the thorium-U-233 cycle and reprocessing, as
described in section 3.12.
These are the typical designs and concepts of ALWRs to improve the fuel
utilization for both once-through and recycling strategies.
Evolutionary Approaches for Development of ALWRs
Some countries continue to adopt large sized units, above 900 MWe, for
ALWRs to be constructed in the 90s, which are proven to be economical and
sophisticated. The French 114 model (1400 MWe) is a continuous improvement
of the P4 series (1300 MWe) and is under construction. It is the latest
generation of PWR in compliance with the French standardization policy and
incorporates the feedback from operating experience (section 3.1).
The Convoy plants (section 3.3) are a group of three plants with PWR of
the standard size for Germany of 1300 MWe, which are presently under
construction. The advanced features of the Convoy concept is mainly in the
field of engineering and project management associated with nuclear power
plant construction as well as the stabilization of the licensing procedure.
The WER-1800 in the USSR (section 3.7), the Mitsubishi-Westinghouse (M-W)
APWR (section 3.9) and Hitachi-Toshiba-GE ABWR (section 3.10) both joint
USA-Japan projects, and the UK Sizewell-B Reactor (section 3.11) (U.S.-U.K.
project) are planned to be constructed in the 90s. They are the designs of
the state of the art incorporating upgrading and advanced technologies in
LWRs.
The other designs with evolutionary approaches described in Chapter 3
offer the diversity of options for the development of ALWRs.
Chapter 4 describes various designs of advanced medium sized reactors
(~ 600 MWe) for the 90s. The designs incorporate a greater degree of
passive safety features, including natural circulation of the reactor
coolant, a gravity driven emergency core cooling system, or passive safety
injection and passive containment cooling etc., as well as more reliable
components and systems and shop préfabrications etc. Laboratory R&D
programmes are being planned, but it is not necessary to construct a
prototype reactor for this approach.
New Concept of ALWR Designs
Chapter 5 describes the conceptual designs of PIUS (process inherent
ultimate safety), including SECURE-P (Sweden), ISER (intrinsically safe and
economical reactor) (Japan) and PIUS BWR and PECOS-BWR (passive emergency
cooling systems for boiling water reactor) (USA). The ECCS (emergency core
cooling system) water supply stored in the prestressed concrete pressure
vessel in SECURE-P is for a cooling period of seven days. In the ISER, the
ECCS water supply, which is stored in a steel reactor pressure vessel, is
reduced to three days. In the PECOS-BWR, ECCS water supply for one day,
further reduces the volume of the steel reactor pressue vessel. The use of
large vessels to contain the reactor core as well as an ECCS water, implies
the possibility of eliminating the pipe breaks and the subsequent loss of
the ECCS water. But further research and development work, including
detailed design studies, as well as construction and operation of a
prototype may be necessary to demonstrate their technical and economic
viability.
LOW-TEMPERATURE NUCLEAR HEAT REACTOR
Nuclear reactors can be used not only for electricity, but also can
supply heat as a primary energy for heating purposes and for industrial
needs. Technical and economic studies in several countries such as USSR and
Canada have shown that the heat delivery from NPPs can be competitive with
fossil-fuelled plants and have a lower impact on the environment. In
principle, all existing types of reactors can be used and some of these are
partly already being used, i.e. PWRs, PHWRs or the Soviet BWR-G (RBMK) and
even the typical BWRs for heat and power co-generation (CHP).
Several countries, like Canada, China, Federal Republic of Germany,
France, Sweden, Switzerland and the USSR, have developed or are developing
specialized nuclear heating plants (NHP). Compared with nuclear
co-generation plants, the specialized nuclear heating plants (NHP) are in an
early stage of development and implementation. There are at least two main
differences in the conception of the heat producing reactors as compared to
reactors of a typical NPP:
a) Due to lower coolant temperature for supply of heat compared to
electricity generation and lower energy demand within the limited
radius of economic transmission, the nuclear heating reactors are
of lower capacity output with lower core power density and with
working pressures about ten times lower than that of typical PWRs.
b) The design of these units incorporates in many cases systems with
passive safety features. Detailed information about nuclear heat
application is given in Refs [12-15]. The summary tables showing
the main characteristics of these NHPs are attached in Annex II.
IAEA PROGRAMME ON ADVANCED LIGHT WATER REACTORS
In order to provide an international forum on the development of the
technology of advanced light water reactors, the Agency has launched in its
Division of Nuclear Power a programme on Advanced Light Water Reactors.
International Working Group on Advanced Technologies for Water Cooled
Reactors (IWGATWR) was established in May 1987. The objectives of IWGATWR
In the areas relevant to advanced technologies in light and heavy water
cooled reactors with emphasis on their safety and reliability:
to assist in defining and carrying out the Agency's programmes in
accordance with its Statute,
b) to promote an exchange of information on national and
multi-national programmes, new developments and experience, to
identify and review problems of importance and to stimulate
co-operation, development and practical application of water cooled
reactors,
c) to provide Member States with information about the current status
and development trends of advanced technologies for water cooled
reactors.
The scope of this Working Group covers:
a) improvements of current water cooled reactors,
b) evolution of water cooled reactor design and technology,
c) new water cooled reactor design concepts.
The focus of the IWGATWR addresses:
programme assessment and planning,
system analyses and fuel utilization strategies,
research, development, design and cost related aspects of
reactor core
plant systems and components
reactor and plant structures and containment,
plant operation and maintenance.
The Working Group will co-ordinate its activities with other Agency
programmes in interfacing areas, as well as with related activities of other
international organizations.
This report represents the first comprehensive effort within the
framework of the IWGATWR to document all major current activities in the
application of advanced technologies to future light water reactors, and
thereby to contribute substantially to meeting the objectives b and c, above.
2. PROGRAMME FOR ALWR DEVELOPMENT
FINLAND [16-19]
In Finland, about 40% of the electricity is generated by nuclear
power. Installed capacity is 2 x 465 MWe PWR and 2 x 735 MWe BWR. Due to
the structure of electric power production and consumption, huge efforts
have been made to achieve the minimum duration of outage and minimum
disturbances during operation. The reloading outages have been 15-30 days.
The load factors of the two Loviisa units have risen to above 86%, and of
the two Olkiluoto units to above 91% (1986).
Activities on advanced technologies for the present light water
reactors are mainly concentrated to measures for core melt accident
mitigation. In Loviisa plant, the new process computer system and the
simulator will replace the old ones and the outside cooling of the
containment shell has been chosen for further studies. In Olkiluoto plant
(TVO I/II), water filling and filtered venting of the containment will be
According to an "Electrical Energy Package Plan" presented in 1986 by
the Ministry of Trade and Industry, some 2700 MWe additional capacity, of
which 500-1000 MWe will be nuclear power, is required by the year 2000.
Before Chernobyl, a new joint company, Perusvoima Oy (PEVO) was founded to
build and operate the next nuclear power units. When feasibility studies
were completed, an application for decision in principle was filed in March
1986 by PEVO. After Chernobyl, the processing of this application was
stopped. However, the public attitude towards nuclear power is recovering
from the Chernobyl-effects. It is expected in the future that nuclear power
will still be considered as the most viable alternative for energy
production in Finland. And the interest will be in LWR development.
Because Finland is not a NSSS-producer, the development work is mostly
concentrated on safety and architect-engineering aspects. In December 1982,
new general safety criteria were issued in Finland. According to these
criteria, core melt accidents have to be taken into account in the design of
new nuclear power plants. Then a severe accident research project was
initiated in 1983. A work to collect the design and safety requirements for
PWR's in the 1990's is in progress. TVO (Imatran Voima Oy) has made in
December 1987 a co-operation agreement with the Swedish ABB ATOM for
development of BWR 90 design adapted to the Finnish conditions. The future
programme for short term targets (to 1991-92) is a guarantee of
licensability according to new requirements, and for the long term (after
about 1995) will present PWR and BWR solutions with further evolutionary
developments.
FRANCE [20-22]
The electricity output generated by nuclear power plants in France is
now 45 GWe (about 70% of total power capacity). By the end of the century,
France may install a capacity of 70 GWe with 60 units, around 90% of the
nation's electrical power. The programme of fossil fuel replacement by
nuclear is coming to the end. The standardization policy which is one of
the reasons for the success of the French nuclear programme compromises
between evolution and stability by continuous evolution in successive series:
900 MWe, 34 units, CPl - CP2 series,
1300 MWe, 20 units, P4 series.
A new subseries of 1400 MWe units has been launched in 1984. The first
unit being due for operation in 1991, and the plants of this model should
still be in operation 40 to 50 years from now.
EOF will keep the same principles for the evolutionary programme which
is proceeded by series, improving the design without significant changes.
For R&D of current reactors, EOF spends 500 Million FF (about US$ 88.5
Million) each year related to improvements of safety, reliability and
availability of operating plants. For example, the M3 project is considered
to increase the power output of the 28 units in the CP1 - CP2 series by 4.3
per cent. For the present PWRs, the lifetime is expected to be extended up
to 40 years. All of the 900 MWe units will progressively evolve toward a
quarter reload scheme, with 3.7% enriched fuel and a 42 000 MWd/t average
discharge burnup. A 10% decrease in the fuel cycle cost is expected as a
result of this improvement.
For future standards of French PWRs in the year 2000 and after, a study
called "REP 2000" ("PWR - year 2000") was started in 1986 by EOF. The
objectives of the standard are: load follow capacity, cost effectiveness
and operation flexibility. With the essential portion of nuclear power, the
grid requires a prescribed load following pattern of the NPP probably with a
new design of control (grey rods). And with the aim of improving the grid,
it is planned to develop an automatic and centralized units power control
systems at the national level. A construction cost reduction of 5% is
expected with the N4 model, but for the period of the years of 2000-2020,
the economic aspects are not really clear at the moment. The conclusions of
the "REP 2000" study should be available in 1988. These studies should be
followed by a preliminary design stage (definition of the main technical
options and of the basic safety options), then by the next design stage in
the beginning of the 1990s. The detailed design and the construction of the
first unit may begin with a commissioning purpose at the beginning of the
next century.
From reprocessed spent fuel, France will obtain a substantial stockpile
of plutonium. In the 1990s, France will enable fabrication of approximately
100 t/yr of mixed uranium/plutonium oxide fuel assemblies (MOX), which were
planned to be loaded in some 900 MWe nuclear plants with a ratio of 307» MOX
and in one 900 MWe plant with a potential 100% MOX fuelled core. Actually,
the first reload is in operation since the autumn of 1987, in the 900 MWe
ST-LAURENT Bl unit with the design burnup level of 33 000 MWd/t. The
recycling program will grow and reach more than 60 t/yr of MOX fuel by 1993.
For further improving of fuel utilization in PWRs, the three French
partners, CEA, EdF and FRAMATOME had jointly started a three year programme
from 1984 - 1987 to assess the feasibility of the convertible spectral shift
reactor RCVS. The FRAMATOME's effort was estimated to be about 40 million
FF per year. Related R&D programmes are concentrating on the validation of
computer codes for core physics and thermohydraulics analysis. An extensive
experimental programme has been undertaken since the end of 1984, with two
critical facilities EOLE and MINERVE at Cadarache and with thermohydraulic
facilities at Grenoble. The first part of the research programme which
defines the feasibility is scheduled for the middle of 1988. It is expected
that no RCVS will be completed before 2002 or 2005 and no detailed design
study should be undertaken before early 90s.
FEDERAL REPUBLIC OF GERMANY [23-25]
In the Federal Republic of Germany around the year 1990, the share of
nuclear energy following completion of plants still under construction will
increase to roughly 40%. Any further expansion will depend on the power
consumption growth rates, replacement of old plants and competition with
coal-fired power plants etc. German experience to date in the construction
of nuclear power plants has consistently confirmed the decrease in power
generation costs as the size of the plant increases. The first 360 MWe PWR
plant commissioned in 1969, which incorporated all of the major features of
future Siemens PWR technology in terms of component and systems engineering
as well as plant design, for example: Incoloy 800 in use for the first time
worldwide as steam generator tube material, reactor coolant pumps with
removable shaft adapter ensuring trouble-free gasket maintenance etc. Only
one year later the 1200 MWe PWR BIBLIS-A plant started construction in 1970,
followed by the orders for eleven 1300 MWe PWRs and BWRs.
At the end of the 1970's, the safety requirements increased rapidly.
The safety philosophy gives the priority to primary safety measures, i.e.
accident-preventing action ahead of measures limiting or mitigating
accidents. The use of very tough materials, as well as low stress levels
and optimized designs mean that the safety of a component is no longer
dependent on stringent fabrication and inspection requirements alone, so the
possibility of sudden failure is precluded. This safety philosophy is
expected in the long run to be internationally accepted. The safety review
of German nuclear power plants confirmed the considerable advantages of the
structure of engineered safety systems and the technology related to
information process and display. Only minor amendments were introduced in
the plants for further reduction of residual risk.
The concept of convoy project processing presented in 1980 envisaged
in-depth standardized planning of the so called power block including
reactor building, reactor auxiliary building, emergency feed building and
turbine building, together making up roughly 80% of a power plant, plus
standard licensing documents. This approach was quite successful, taking
into account Germany's federal structure and consequent decentralization of
responsibilities and procedures. The three projects being processed along
these lines are below the original budgets. The entire construction
activities being covered by two construction licenses and one commission
license in each case. Time schedules could have been shortened.
The further development of Advanced LWR with its large potential will
place emphasis on shortening construction periods, reducing costs,
automation of plant operation and perfection of service activities with the
aid of specialized tools and procedures. Intelligent computer systems will
take over supervision and control of the entire plant. Recently Siemens is
considering a 1000 MWe 3 Loop PWR for a number of international proposed
sites. The new plant follows closely internationally applied safety and
licensing practices. The improvements mainly are optimization and
simplification of the design in the sense of improving operability,
maintainability and economic viability of the entire plant.
The Federal Republic of Germany is also interested in the recycling of
reprocessed plutonium and residual uranium. As of March 1987, more than
25 000 Mixed oxide (MOX) fuel rods had been inserted and irradiated in PWRs
and BWRs, the maximum burnup achieved being beyond 52 000 MWd/tHM. A joint
development of a high converter design basis by Kernforschungszentrum,
Karlsruhe, Swiss Federal Institute for Reactor Research (EIR), at
Würenlingen, Technical University of Braunschweig and Siemens is going on.
With a high conversion ratio (~ 0.9), reduction in fuel consumption of
50-70% would be obtained, compared with the once-through fuel cycle at
current PWRs, without having to alter major existing PWR technologies.
Relevant integral measurements are carrying on in zero-power reactor
facility, PROTEUS at EIR. A 5 MW high pressure water test facility has been
installed at the Siemens Karlstein research centre for the
thermal-hydraulics.
Detailed design studies related to the core and pressure vessel internals
for a light water high converter are under way at Siemens. In order to keep
the fuel cycle costs acceptably low, it is necessary to aim at high
discharge burnup and longer cycle length. In-pile tests with steel clad
subassemblies are planned at the Obrigheim PWR.
2.4 JAPAN [26-28]
In Japan, nuclear power plants now in operation total 36 units for
28.046 GW (October 1987). The installed nuclear capacity would increase to
34 GW (about 19% of total installed capacity) by 1990, 48 GW (23%) by 1995
and 62 GW (27%) by the year 2000. The targets for further development of
LWR are sophistication over three generations of LWRs: existing plants (now
in operation and under planning for operation in the mid-1990s), advanced
LWR plants, and the next generation type of LWR plant.
The technical development of APWR (Mitsubishi-Westinghouse) and ABWR
(Hitachi-Toshiba-General Electric) are well under way. MHI (Mitsubishi
Heavy Industries, Ltd.) and Westinghouse have been working together under
the support and guidance of 5 Japanese utilities in an Advanced PWR
programme since 1981. In the design, a moderator control feature has been
added, using water displacer rods referred to as mechanical spectral shift.
By virtue of fuel cycle prolongation and other improvements in APWR, the
fuel cycle cost can be reduced about 20% as compared with current PWR's.
Significant reductions in the requirements for both separative work units
and uranium are accomplished. Verification testing of the major components
was completed by early 1987. An extensive review programme of the design
has taken place with U.S., Belgian, and Japanese utilities. Construction
site has not been decided yet.
ABWR development began in 1978 with the formation of the Advanced
Engineering Team (AET). Organized by General Electric, AET consisted of
technical specialists from the worldwide BWR suppliers - Ansaldo Mecanonico
Nucleare SpA (AMN) of Italy, ABB-ATOM of Sweden, General Electric, plus
Hitachi Ltd and the Toshiba Corporation of Japan. During 1978 to 1979,
referred to as Phase 1, AET developed a feasible conceptual design of an
imprpoved BWR. Phase II of ABWR development was an integral part of the LWR
Third Improvement and Standardization Programme undertaken by the Japanese
Government, utilities, and manufacturers. During Phase II, General
Electric, Hitachi and Toshiba engineered a detailed design which was
evaluated favourably, and conducted a wide range of tests to confirm the
reliability and performance of the new technologies to be employed. Phase
III, the final phase in the ABWR's development, came to a close in December
1985. The purpose of this was to simplify systematically the ABWR and
reduce its cost.
The ABWR is now ready for lead project application in Japan. The ABWR
has been selected by the Tokyo Electric Power Company for its next two units
at the Kashiwazaki-Kariwa Nuclear Power Station site. These units are
planned to begin commercial operation in July 1996 for K-6 and in July 1998
for K-7. (K-6 and K-7 stand for Kashiwazaki-Kariwa Unit 6 and 7
The basic direction of the next generation of LWRs - Japan (called
AA-LWRs) will be designed by modification of A-LWRs to meet future social
and economic requirements. The technical development of the next generation
of LWRs will take more than 10 years, and a further 10 years until the
installation of the first unit, which is expected to he around the year
2005. In order to meet future social and economic needs, the next
generation of LWRs will be aimed at further enhancing the functions of
reactor cores, enhancing fuel performance, improving safety design
technique, utilizing more advanced technologies, and improving aseismic
technology (siting free). The highlights of the developmental targets are
listed in Table 2.4.1.
TABLE 2.4.1
DEVELOPMENTAL TARGETS FOR SOPHISTICATION OF LWR TECHNOLOGIES
Existing LWRs
(results in FY 198A)
Sophistication of
(Continuous Operation
Time: 11 months, Periodical Inspection
(15 months)
(60 d)
Development of
A-LWRs
AA-LWRs
10% Reduction in
kWh Cost from
(Over 15 months)
(50-60 d)
Time: (80-120 d)
Reduction of
Exposure Dose
LLW Products
Expansion of
3.7 manSv/
reactor-yr
1600 drums/
Limit to feasible
siting places
2/3 of
current average
10-20% from
Over 10% from
0.5-1 manSv/
Less 0.5 manSv/
100-200 drums/
Less 100 drums/
To expand
feasible siting
land by siting
on quaternary
period layer and
adoption of
earthquake isolation design
For the utilization of plutonium in light water reactors, a three-stage
plan has been prepared by MITI (the Ministry of International Trade and
Industry). The plan includes a small scale test programme scheduled for the
immediate future, a large scale demonstration in the first half of the
1990s, and a full scale use scheduled for the second half of that decade.
The small scale demonstration programme was designed to use two MOX fuel
(uranium-plutonium mixed oxide fuel) assemblies in the Tsuruga Power Plant
Unit 1 (357 MWe BWR) in 1986, and four MOX fuel assemblies in Mihama Power
Plant Unit 1 (340 MWe PWR). For large-scale demonstration, one BWR and one
PWR will each be loaded with MOX fuel assemblies up to one quarter of the
core. First loading in BWR is planned around 1992 and in PWR around 1994.
The power rating of the BWR and the PWR will be at least 800 MWe output.
For the full-scale demonstration programme and the beginning of commercial
use with both the BWRs and PWRs the start time will be around 1997. For the
utilization of recovered uranium from the reprocessing of spent fuel by some
time around 1995, specific studies are planned considering re-enrichment
process for use in LWRs and as material for MOX fuel.
In order to assure the promotion of nuclear power for 21st Century, it
is necessary to perfect safety assurance measures. A 'Safety 21 Committee'
was organized by MITI in March 1987 to determine 'Safety 21: Improvement of
Safety Assurance Measures for Nuclear Power Generation', to perform safety
plan steadily and to continue the efforts for improvement of safety. In
April 1987 a 'LWR's Technology Sophistication Committee was organized to
review the total research programme for LWR developments. Under the
Committee there are 5 groups:
Next Generation LWR's Working Group
Existing and Advanced LWR's Working Group
Fuel Technology Working Group
High Technology for Seismic and siting Working Group
Investigation of Foreign Technologies
In Sweden, 12 nuclear power plants are in operation, of which 3 are
PWRs and 9 are BWRs, and they produce 50% of the total electricity
generation. The operating experience of the nuclear plants has been good
with high capacity factors and low occupational radiation exposure figures.
In 1986, the average capacity factor was 80.5%
(85.4% for the BWRs).
annual occupational radiation exposure has throughout the years been around
1 manSv (100 man rem) per reactor unit (for the BWR plants).
However after a referendum in 1980, which resulted in a majority for
completion of the 12 reactor programme, the politicians decided that then no
more nuclear plants were to be built, and that the energy policy should aim
at a total phase-out of nuclear by the year 2010.
In 1981, the Government stipulated that means to control and minimize
release of radioactive matter to the environment in the event of one extreme
accident (resulting in a degraded core) were to be provided at all the
operating nuclear power plants before 1989. Consequently, the utilities
have concentrated their development efforts to the plants in operation,
partly to meet the "degraded core accident" requirment, partly to improve
the operational flexibility and reliability of the plants, and to simplify
The FILTRA installations (filtered vented
containment systems) are now nearing completion at the different plants.
The utilities have also been engaged heavily in the development
activities related to the back end of the fuel cycle, and to a safe final
storage of low and intermediate level radioactive waste, for which the first
stage of the repository has been taken into opération at the Forsmark site,
under the sea bed of the Baltic. As for the fuel cycle back end, the
preferred solution was to build an intermediate storage facility CLAB (at
the Oskarshamn site) where the spent fuel from all the nuclear plants is
stored for a period of 35-40 years. Then the spent fuel will be "packed" in
canisters and placed in a final repository, a tunnel system in stable
bedrock at a level of about 500 m below grade. The construction of this
repository will probably not start before the turn of the century, but quite
a lot of development work is being carried out in various areas.
With no near term prospects of new nuclear plants to be built, ABB
ATOM, the only nuclear plant vendor in Sweden, is directing a major portion
of its development activities to support the utilities and the plants in
operation, but is also actively pursuing various research and development
programmes independently. Examples are: water chemistry and material
properties programme to eliminate IGSCC (inter-granular stress corrosion
cracking), and to minimize radiation levels for maintenance, wet oxidization
of radioactive waste and solidification in cement, improved computer systems
and special computer programmes for improved monitoring of plant and fuel
conditions, and fuel development - advanced BA (burnable absorber)
strategies for BWR and PWR fuel, reduced susceptibility to PCI (pellet clad
interaction), and for increased burnup. A new generation of SVEA fuel with
10 x 10 array and 9 mm rods is introduced instead of 8 x 8 array and 11 mm
rods.
ABB ATOM has made a thorough review of its BWR design, based on the
experience from the construction, commissioning and operation of the
Forsmark 3 and Oskarshamn 3 plants, with the aim of evaluating possible
improvements, simplifications and cost reductions. The result, the BWR 90,
has significantly reduced building volumes, shortened construction time, and
decreased amounts of systems and components, and includes measures for
simplified operation, testing and maintenance, i.e. the costs will be
lowered, and the plant operation more simple. Design measures to cope with
a "degraded core" accident have also been included in the new concept, so
that the public and the environment should be protected even in such a low
probability event.
ABB ATOM has furthermore for more than a decade been working on reactor
design concepts, the SECURE reactors, in which the nuclear safety is based
on simple immutable natural laws (gravity & thermo-hydraulics) only - the
PIUS principle. In LWR, major release of radioactive matter to the
environment with prompt or delayed significant health effects must always be
preceded by core overheating or melting (core degradation), and prevention
of core degradation therefore guarantees absence of serious accidents. In
the SECURE reactors, the PIUS design principle is applied uncompromisingly
to ensure protection of the core against overheating and melting under any
credible or conceivable conditions. This means that the "degraded core"
accidents are completely eliminated, i.e. in the SECURE reactors the utility
will be protected against the risk, even though very low in other LWR
plants, of a core melt and its disastrous economic consequences. The PIUS
design study of a 600 MWe unit with external pump and steam generation loops
is proceeding.
UNITED KINGDOM [30-31]
In the United Kingdom, it is foreseeable that a strong nuclear industry
and a sizeable nuclear power generation are needed, because the conventional
resources are difficult to meet the country's long term energy needs. A
very extensive public inquiry for the Sizewell B nuclear power station was
set up by the Government, lasting from January 1983 to March 1985. This
inquiry examines very wideranging terms of reference, including need,
economics, site and local environmental consequences, as well as safety,
which will have a major long term effect on safety assessment attitudes in
the UK. After that, the British government gave full planning approval and
full financial approval to the construction of a PWR at Sizewell in Suffolk
in March, 1987. Work on the site began soon afterward. Before the end of
1987, the Central Electricity Generating Board (CEGB) applied for planning
approval for a PWR at Hinkley Point and the further PWRs in subsequent years.
Over 90% of the Sizewell-B project will be spent in the U.K., the value
of imported hardware being only 3%. Contracts have already been signed with
British companies for the steam generators, turbine generators, pressurized
and high pressure pipework as well as reactor coolant pumps and motors. The
assessments made would allow the Sizewell-B to provide electricity at a
substantially lower price than the AGR, Advanced Gas (C02> Cooled
Reactor. A unit cost of 2.33p/kWh would be achieved against a projected
3.05p/kWh for a new AGR. These lower costs stemmed from the PWR's lower
capital cost, lower fuel costs and longer life (up to 60 years for ALWRs).
UNION OF SOVIET SOCIALIST REPUBLICS [13,32]
At the end of 1986, the Soviet Union had 27.7 GWe of capacity in
operation, accounting for about 10% of national electricity output. Over
50% of this capacity comprises RBMK (graphite-moderated boiling water,
pressure tube reactor). Because of the Chernobyl accident, the number of
RBMK plants in the USSR will be limited to 21, including 13 in operation and
8 in various phases of construction. Future nuclear power plants will be of
the WER type. The installed nuclear capacity is projected to be 100 GWe by
the year 2000, about a 25% share of USSR electricity generation.
Further development of nuclear power in the USSR will be evolutionary,
based on the experience accumulated from plant operation and design. 10
units of the WER-1000 MWe were designed between 1974 and 1980. For
improvement on these units, the design is updated with certain features, for
example: the increased use of the passive systems, the installation of
diagnostic and automatic process control systems and the increase of
effectiveness of the control mechanism, etc. The design of the WER-1800,
(5250-5800 MWt), an upgraded, more economical and safer version of the
WER-1000 PWR, has been started by the Kurchatov Institute of Atomic
Energy. It is planned to build two WER-1800 by the year 2000.
In the USSR, the electricity and heat co-generation (CHP) technology is
Use of
special turbines at such CHP-NPPs permits the heat supply from one 1000 MWe
power unit to be increased up to 900 Gcal/h. By the year 1990, the
construction of this kind of CHP-NPPs is envisaged in several cities. A
special heating reactor, AST-500 (500 MWt) has been developed, designed for
generation of heat in the form of hot water. AST-500 is a water-water
vessel-type reactor with respect to the inherent safety principles used in
the design, which can be built in a close vicinity of the population
center. The two AST-500 pilot plants now are under construction.
used at several NPPs in different sizes of PWRs as well as RBMK.
UNITED STATES OF AMERICA [3,33,34]
In the United States, the size of the U.S. grid is approaching 700 GW
of generating capacity. If demand for electricity were to grow at a rate of
3% per year, the U.S. would need to add some 21 GW of new capacity each
year, after present excess capacity has been absorbed. This is equivalent
to 21 large, 1000 MWe generating stations. If a significant fraction of
this growth is met by nuclear power plants, this would be by far the largest
potential market in the world.
The United States has embarked upon an aggressive LWR revitalization
programme. The Department of Energy and the Electric Power Research
Insitute (EPRI) have jointly developed an advanced light water programme to
be pursued over a five-year period. A key feature of this programme is that
it joins elements of the utilities (EPRI), NSSS vendors, architect
engineers, Nuclear Regulatory Commission, and the Department of Energy into
a single comprehensive effort to design and certify the next generation of
LWRs for utilization by the U.S. utilities.
The programme has four major parts:
to determine a set of stable regulatory requirements which must be
met by the next generation LWR,
2. to generate a utility and NRC approved plant Requirements Document
for the next generation LWR nuclear power plants,
3. to produce the detailed engineering designs and obtain NRC
licensing certification for the next generation LWRs, and
4. to produce designs for a medium-capacity APWR and ABWR.
As a first step in establishing design requirements for the next
generation of LWRs, it was necessary to work with the NRC to determine the
safety and licensing requirements which would have to be met by the new
design. The purpose of the Utility Requirements Document is to use utility
operating experience with current generation plants to generate the design
specifications for the next generation. The new reactor designs generated
under this effort are evolutionary rather than developmental in nature; that
is, they rely on proven design concepts insofar as possible. No design which
would require extensive development and building of a prototype reactor for
concept demonstration is being accepted. Emphasis is being placed on:
elimination of unnecessary complexity
evaluation of design margins
improved operability and maintainability; and
improved constructibility
In 1991, the GE and CE large plant designs are expected to get final
design approval and certification from the NRC. Westinghouse is also
pursuing NRC approval of a large plant design via a program initiated prior
to the DOE/EPRI program.
In the final part of the programme, DOE and EPRI have an ongoing effort
on the design of medium-capacity(600 MWe) LWRs. Phase I of this programme,
to generate initial conceptual designs, was completed. Various design
options were screened, and the most promising conceptual designs were
selected. Phase II, which involves further work on the most promising
designs and co-operative design and testing supported by the U.S. Department
of Energy, began in June 1986, and will continue through 1989. For the 1988
fiscal year, Congress has appropriated US$ 17 million for DOE to use in
support of the entire ALWR programme. When the Phase II effort is completed
at the end of 1989, the completed conceptual designs will be evaluated by
industry and government.
In 1989, the development of comprehensive requirements and criteria
which will improve construction quality, cost and duration will be
completed, the improvement of instrumentation and control systems will be
achieved and the conceptual design development and required testing for
Westinghouse and GE mid-size plant designs will be accomplished.
For uranium utilization improvement, the U.S.A. concentrates on the
once-through fuel cycle, because spent fuel is not expected to be
reprocessed in the foreseeable future. The greatest part of this interest
has been focused on backfittable improvements, i.e., on those improvements
that can readily and economically be implemented in both existing and future
plants [34].
CMEA Member Countries [32,35]
At present, over 20 nuclear power plants with a total capacity of about
36 GWe, operate in the CMEA member countries:
Bulgaria, Hungary, the GDR,
the USSR and Czechoslovakia, in 1986, accounting for 9.9% of total
electricity generation. Over 80 units of nuclear power plants and nuclear
heat and electricity plants (NHEP), with about 70 GWh capacity, are under
construction or preparation for construction, in Bulgaria, Hungary, the GDR,
the Republic of Cuba, Poland and Czechoslovakia and the USSR. In the
comprehensive Programme of Scientific and Technological Progress up to the
year 2000, the accelerated development of nuclear power is incorporated as
the third priority area. The CMEA member countries are convinced of the
necessity to develop nuclear power at a higher rate, as compared with the
traditional energies. Under the Programme in the CMEA member countries,
except the USSR, by the year 2000, the electricity generation by nuclear
power will increase from 8 GWe in 1986 to 50 GWe, about 30-40% of the total
electricity generation. The implementation of such an extensive programme
needs close co-operation in science, technology and production. An
agreement was signed on the multilateral international specialization and
co-production and mutual deliveries of equipment for nuclear power plants.
The industry of eight participating countries (Bulgaria, Hungary, the GDR,
Poland, Romania, the USSR, Czechoslovakia and Yugoslavia) specializes in the
production of equipment, fittings and instruments with WER-440 and
WER-1000 reactors.
The Soviet Union produces virtually all kinds of equipment, and
supplies the CMEA member countries with over 50 per cent of basic equipment
for WER-440 and WER-1000 units.
In the CSSR, the nuclear power share of total energy consumption will
increase to about 17% in the year 2000 as compared with 13.7% in 1980.
Preparation for the construction of 4 units with WER-1000 reactors has been
under way. The CSSR ranks second following the USSR as the biggest
manufacturer and supplier. Specialized production and supplies from the
CSSR cover 80 per cent of a whole range of technological equipment for
WER-440 and WER-1000, including reactors, main shut-off valves, separator
of steam superheaters, special pumps, fittings, as well as control and
instruments.
In the German Democratic Republic, the future electricity requirements
will be met solely by expanding nuclear power capacity, all by WER-1000.
The two units are due in service in 1991 and 1993.
In Poland, the construction of the first nuclear station with 4 x 440
MWe WERs began in 1985. The units are due in service between 1990 and 1994.
In Bulgaria, two WER 1000 reactors are under construction. A 1000 MWe
WER is planned to be added to the grid every other year. Nuclear power
will share 40% of total electricity generation in 1990 and 60% in 2000.
After the Chemobyl accident, the prospects for nuclear power was not
struck out in the CMEA member countries. But the research and development
are more focused on safety and reliability. Within the framework of the
Comprehensive Programme, the co-operation on the improvement of safety for
current reactors includes the development of new types of equipment,
non-destructive inspection and diagnostic instruments, the automazation of
the control system, reconstruction and modernization of operating plants, as
well as the elaboration of safety regulations, etc.
2.10 DEVELOPING COUNTRIES [36-42]
There were 24 nuclear power reactors with a total capacity of ~ 14
Gwe in operation in seven developing countries including: Argentina, Brazil,
India, Republic of Korea, Pakistan, Yugoslavia and Taiwan, China (1986).
There are 15 units under construction, with a total capacity of 9.5 GWe in
countries including China, Cuba, the Islamic Republic of Iran, and Mexico
etc. It is clear that in some developing countries there is a need for the
development of nuclear power. But because of various reasons and
constraints, i.e., insufficient trained manpower, inadequate infrastructures
and economic and financial problems, the increase of nuclear power over the
next 15 years still will be limited. The forecast shows that nuclear power
capacity is expected in developing countries to rise to 36 GMe, only a 5.7%
nuclear share of their electricity generation by the year 2000.
The development of nuclear power in Argentina and India will be still
along the line of heavy water reactors incorporating advanced technologies.
In particular, India is using the fuel cycle with thorium. But some efforts
in Argentina point to advanced LWR technologies:
With UNDP-IAEA support, an academically inspired research program
in the field of tight lattice cores with high conversion ratio has
been started: work is in progress in resonance treatment,
homogenization models for cell calculations, and validation of
nuclear data and codes.
The basic engineering design and a preliminary economic feasibility
study have been completed for CAREH, a 15 MWe modular LWR with
minimum on-site installation work, passive emergency systems and
automatic operation. Its primary circuit is integrated, self
pressurized and natural convection driven. The development of
different CAREH subsystems is under way.
In Korea there are now 6 PWRs (2 x 600 MWe, 4 x 950 MWe) and 1 PHWR
(600 MWe) in operation that share over 50% of the total electricity
generation, and 2 x 950 MWe PWRs are under construction, while another
2 x 950 MWe PWRs are in the design stage. The three 600 MWe plants as the
first phase of nuclear power projects were constructed under the turn-key
basis in the early 1970s. The ratio of domestic supply to the total
project, which is usually called as a localization factor, in this phase was
only about 8 to 14 per cent. Some domestic companies have participated in
construction works as sub-contractors to foreign suppliers. Six 950 MWe
plants, as the second phase of the project, have been carried out under the
framework of the non-turn-key contract or the component approach. The
localization rates reached about 45 per cent, mostly in the fields of
architect engineering technology and equipment manufacturing technology. In
order to effectively achieve self-reliance in nuclear power technology by
the year 2000, the design of 2 x 950 MWe PWR nuclear steam supply system
(NSSS) has been assigned to KAERI (Korea Atomic Energy Research Institute)
with a foreign NSSS supplier. The Korea Nuclear Fuel Company (KNFC) is
responsible for manufacturing and supplying all PWR fuel from 1989.
In Brazil, the nuclear programme is based on the full technology
transfer within a long-term association with the partner of the Federal
Republic of Germany. The means of technology transfer include the joint
design and construction of a series of identical plants, the establishment
of joint enterprises to execute this programme, training, transfer of the
full technical information etc. As a result of this strategy, the
construction of the first 2 x 1300 MWe Siemens type PWR, Angra 2 and 3 were
started. Now 70% of the civil works of Angra 2 are concluded and the
preparatory site works of Angra 3 are ready. And 70% of the design is
completed. However, due to the economic crisis in the 1980s, the programmme
had to be slowed-down. The following plants, Iguape 1 and 2, on which work
was started, were postponed indefinitely in 1983. The restart of
construction of additional NPPs is not expected until the next decade.
China, through many year's efforts, has built up to some extent a
nuclear industry and man-power resources. Long before the prototype 300 MWe
Qinshan nuclear power plant was started to be constructed in 1985, its R&D
programme had been carrying on since 1974. The programme consists of main
items which are necessary for confirming and verifying the design, including
reactor core physics, thermal-hydraulics, materials, fuel, and structure
mechanism, etc. The capacity of manufacturing includes the following main
components: steam generator, pressurizer, reactor internals and fuel
assemblies. Meanwhile, foreign companies were invited to conduct
consultations and some important equipment, such as the reactor pressure
vessel, coolant pump were imported. The first unit of the 2 x 900 MWe Daya
Bay power plant started construction in August 1987. It is planned to be
completed in 1992, and in 1993 the second unit. The NSSS and Turbine
generators are supplied by Framatome and GEC, respectively. Now at the
Qinshan site it is planned to construct two 600 MWe plants. For this
project, it is planned to perform design, component manufacturing, and
construction domestically with incorporation of technology transferred from
foreign partners. In Taiwan, China, it is being planned to place new orders
for LWR power plants. In Southwest Center for Reactors (SWCR) a conceptural
design of AC-600 (600 MWe Advanced PWR) was started in 1987, incorporating
advanced reactor core, passive safety systems and simplification. In
Beijing Nuclear Engineering Institute a feasibility study of SECURE-H for
district heating of Qigihar City, Northeast China has been carried on
together with the experts from ABB-ATOM. China will take a spent fuel
reprocessing option. Since the middle of the 1970s, the R&D of reprocessing
technology has been carried on. The conceptual design of a multi-purpose
reprocessing pilot plant is under way.
3. LARGE SIZE ALWR DESIGNS (above 600 MWe)
THE N4 MODEL [43] (FRANCE)
The N4 model (1400 MWe), which is under construction on the CHOOZ site,
is a continuous improvement of the P4 series (1300 MWe) and is the last
generation of PWR in France (commercial operation in 1991). The plant cost
per installed kW is 5% reduction compared to the P4 series. This reduction
results mainly from the following evolutions.
Fuel Utilization and Core Configuration
The core power is increased by means of loading 205 Advanced Fuel
Assemblies (AFA) instead of 193 fuel assemblies in a 1300 MWe plant. For
Advanced Fuel Assemblies, the grids are made of Zircaloy 4 instead of
Inconel 718. This allows a gain of about 0.04% on fuel enrichment and
reduction of stagnant activity in the primary system. The top and bottom
nozzles of the assembly can be removed to replace a failed rod.
For fuel economy, burnable absorbers and neutron reflector etc. are
considered. For fuel management, the reloading pattern will be as for the
other reactors of EOF by 1/4 of the core taking into account the increase of
the burnup (more than 45 000 MWd/t), and the interest for EOF to keep the
annual loading, because of the network consumption (low in the summer).
Load Follow Capability
For increasing load follow capability, a new control system called DMAX
is selected. It involves five control rod banks (two grey banks and three
black banks). The overlap length between two banks is controlled by the
system, which simultaneously controls the average primary coolant
temperature and the core axial offset. In this way, the operator action is
no longer required during power transients. The control banks can
compensate the reactivity variations of the xenon, therefore reducing
radwaste during power transients. The position of the grey bank is no
longer controlled by the turbine power signal, so reducing the interface
between turbine and NSSS. The DMAX system allows automatic follow-up of
load variation requests up to 5% of maximum power per minute. The new units
have to enable frequency adjustment involving primary adjustment and
automatic frequency control. The units will participate in the spinning
reserve, when automatic operations are inadequate using available power
reserves to support the grid. These load followings can be performed up to
70% of the fuel cycle length (boron control system design). The number of
load following transients allowed in the unit life is 12 000.
The new control room is equipped with the wide-scale integrated control
and data display facilities, a high performance data processing system,
which enables excellent coordination between the operation, the maintenance
and periodic tests. In the control room, the control, in all circumstances,
would be only achieved by computerized control and control display units.
Conventional equipment would be limited to emergency controls (emergency
shutdown, backup procedures). Programmable and self-testable controllers
are used to receive, process and transmit on, off and analog signals. A
multiplexed optical fiber communication network significantly decreases the
number of cables.
The control and instrumentation system is provided with a two level
structure. The first level provides the interface with the plant (data
acquisition, command of actuators). By itself it carries out all the
automatic protection actions and all the automatic regulation. This level
animates the wall-mounted mimic panel and directly receives, from the
control room, the manual orders for safeguard of protection actions and the
command signals from the auxiliary control panel. All other commands
(on/off and adjustments) are transmitted from the operator consoles to the
first level via the second level. The second level also receives nearly all
the data gathered by the first level. This second level, the "manager" of
the operator consoles, supports all the dialogues, conducts elaborate
processing on the partially processed data coming from the first level, and
executes all the operator assistance and filing functions.
There are two identical, general purpose control stations, from which
two seated operators can perform all control functions in all
circumstances. One operator only is required in stable or low-disturbance
conditions. From three visual display units, the operator can find out, at
any time, the status and configuration on the centralized power plant
control systems and the values of the main physical paramaters, can control
the actuators and can select control procedures applying to the present
situation of the unit and to the concerned system. The alarms, after being
initially processed by the computer system, are displayed on four
specialized visual display units. Then their processing is performed by
means of three control visual display units.
One auxiliary station, known as the "observation station", allows other
users to access the same information as the operators, but without accessing
the controls. On a large passive mimic board, the main loops of the unit
are represented with the main parameters in a highly simplified way. An
auxiliary panel with conventional control and data display devices is
provided in case of short-term failure of all the data processing units. In
a room located near the control room, there is a fourth "observation-only"
station designed for operating, test and maintenance personnel and, in times
of crisis, for the local crisis team. The control and instrumentation
system is shown in Fig.3.1.1.
PWR 1400 MW UNITS
CONTROL AND INSTRUMENTATION SYSTEM
MIAMI M/WfC
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FIG.3.1.1. Control and instrumentation system.
3.1.4 Nuclear Steam Supply System and Turbine Generator
Reactor Vessel
The N4 model vessel is manufactured from hollow ingots which improves
the material quality. The characteristics of the vessel shell material are
improved reducing the initial transition temperature and decreasing
impurities in Cu and P content, in order to increase margins at the end of
plant life. The inside diameter of the vessel penetration tubes is reduced.
Steam Generator (Fig.3.1.2)
Integration of an axial flow economizer provided with a
feedwater-recirculating water mixing system,
use of Inconel 690 as a tube material with the aim of reducing
corrosion risks as well as nickel and cobalt looseness,
selection of a tube bundle with triangular pitch which allows
decreasing its volume while increasing the exchange surface
(7300 m2 instead of 6900 m2 for P4),
modification of the moisture separator design in order to make them
more compact,
measures making maintenance operations easier (Hanway diameter
raised to 450 mm, easier tube plate cleaning),
layers tubing to avoid the problems of perpendicular wear on the
anti-vibration bars.
TUBULURE SORTIE VAPEUR
UMITCUn OE OÉ1IT
8LOC SECHEURS
MOOULAIBES
TROU D'HOMME
BLOC SÉPARATEURS
•CYCLONE*
ENCEINTE SECONDAIRE
FAISCEAU TUiUUURE
JUPE ENVELOP«
DC rAISCCAU
DOUflLIJUPC
PLAQUE SUPPORT
CLOISON ECONOMIWUN
ANNEAU SUP«*!
PLAOUÏ DC DISTRIBUTION
PLAQUE TUBULAIRE
PLAQUE OC SEPARATION
TuwuM OE SORTIE PRIMA«
BOlTE A EAU PRIMAIRE
FIG.3.1.2. Steam generator.
These improvements result in a steam pressure increase of 72 to 73.3
bar going along with a reduction of weight and space required.
Primary Pump (Fig.3.1.3)
VOLAKT
ARBRE MOTEUR
DEMI-ACCOUPLEMENT
DEW-ACCOUPUEMEKT
AABRE POMPE
JOINT M" 3
PAUER POMPE
GUIDE D'ASPIRATION
FIG.3.1.3. Primary moto-pump group.
Primary pump has following improvements:
improvement of the pump compactness and efficiency ( + 2%) by means
of a new hydraulic design (redesigned casing),
use of a hydrostatic bearing, mounted around the impeller, which
eliminates the pump shaft overhanging position, and thus reduces
vibrations perpendicular to the shaft sealing joints,
new oil pressure shaft coupling system,
a thermal barrier with cooling coils of radial design,
modification of the grade of the metal used for the shaft.
The new ARABELLE turbine was specially developed for N4 units.
It differs from the previous 1300 MW turbines by its compactness
(reduced length and weight of respectively 10 and 15%),
greater efficiency
(+ 1%) and increased power (steam admission 8650 t/h at 71 bar instead of
7780 t/h at 69.5 bar).
The major technical innovations are the following ones:
The turbine is of the impulse type, whereas those of the 1300 MW
series were of the reaction type.
The use of a single-flow HP-MP cylinder, instead of the single
HP cylinder, involves a reduction of the dimensions of the three
LP cylinders (the number of stages in these cylinders is divided by
two) .
The outer shell of the LP casings and the cylinders of the turbines
are completely independent; the shell is an extension of the
condenser directly resting on the ground; it is connected to the
cylinder by two flexible circular seals. As a result, the turbine
is subject to no permanent or transient load by the condenser.
The adoption of two bearings for each LP cylinder reduces the shaft
vibration level.
For higher performance, two two-stage moisture separator reheaters
increase the efficiency of the turbine (+ 0.4%).
Balance of Plant
Nuclear auxiliary building: a complete new design linked to the new
boron recycle system.
Fuel storage building: a new fuel storage pit liner design allowing
radiography of welds.
Site radwaste building: a new design - all the site buildings
presenting a potential contamination risk outside Nuclear Island have
been gathered in this building (hot warehouses, hot workshops,
washhouse, liquid and solid radwaste storage and treatment).
Boron recycle system: a new design without intermediate storage
tanks. A boric acid evaporator is used for gas stripping.
Auxiliary feedwater system: two independent subsystems. Each
subsystem includes a turbine-driven pump and an electric engine pump.
Chemical and volume control system: a throttling valve instead of
letdown orifices.
Condenser circulating water system: the natural draft cooling tower
efficiency has been improved by a gutter system that catches droplets
before they can reach the ground level and by a more efficient design
of air inlets.
Ventilation systems - are simpler than in previous projects.
Maintenance Conditions
Development of New Maintenance Equipment
quickly set up closure plates for channel head of steam generators
(light materials, foldable plates),
wider steam generator manholes (450 mm instead of 400),
steam generator cleaning equipment (slurry flushing),
new multiple studs tensioner,
fixed and mobile devices for in-service inspection.
Reduced Exposure of Operators
implementation of a cold purification pump (reduction of the manSv
inventory by 6%),
choice of equipment with low activable product release rates
(Inconel 690 for S.G. tubes, Zircaloy 4 for fuel grids, valve
coatings without cobalt),
improved surface conditions (pool coatings, S.G. channel heads),
improved maintenance conditions owing to the development of
high-performance tools and robots,
improved building design and layout (for example: neutron
containment in the reactor pit, Fig.3.1.4).
IWMIB mini IVLJJII • i ii
••—•)'
Ï..HOLD-DOWN SPRINGT:
IflEACTOR VESSEL
<•
FIG.3.1.4. Chooz-B1 PWR reactor cavity.
3.1.7 Safety related improvements
improvement of redundancy on the main components of the SG
auxiliary feedwater supply systems and of the components cooling
systems, since those systems are safety related and are frequently
or permanently used;
deeper knowledge of the physical phenomena occuring under incident
or accident conditions associated with probabilistic studies in
order to check the whole safety systems redundancy, reliability and
coherence ;
improvement of the man-machine interface (control room);
development of the main components in-service inspection methods.
CONVERTIBLE SPECTRAL SHIFT REACTOR (RCVS) [44-47] (FRANCE, FRAMATOME)
3.2.1 RCVS Concept
In the convertible spectral shift reactor (abbreviated "RCVS", for its
French name) concept, the greatest flexibility of fissile material use has
been sought. Thus such a reactor would be able to use not only uranium
fuel, like conventional PWRs, but also plutonium fuel or mixed uranium and
plutonium oxide (MOX) fuel. For a plutonium-fueled RCVS, the spectrum
(Fig.3.2.1) designed favours plutonium 241 fissions and leads to an
efficient buildup of this isotope, via the 1 eV resonance strong absorption
of Pu-240. The plutonium 241, having a half-life of 14.7 years, is
transformed into americium Am-241. Am-241 captures free neutrons to become
Am-242, which, in view of the spectrum, is also an excellent fissile
100keV
1 keV
1 0 e V 0.1 eV
0.001 eV
1 MeV | 10keV l
100 eV i
1 eV i 0.01 eV
FIG.3.2.1. Absorption rate tilt produced by control rod insertion in the RCVS core.
Studies have confirmed that if the lattice pitch is reduced to give a
moderator-to-fuel ratio of 0.6, it is then possible to achieve a conversion
ratio greater than unity using plutonium fuel. But this type of lattice
requires 8 to 9% high enrichment of fissile plutonium, so raising fuel
costs. Safety studies further demonstrate that enrichment beyond a maximum
of 7.5% of fissile plutonium presents a risk of the reactor returning to
criticality if the core is uncovered.
For the near-term goal, it is essential to use the same facilities for
fuel fabrication and reprocessing as used for current PWRs. This constraint
leads to the choice of the moderator-to-fuel ratio of 1.1, which implies a
fairly low conversion ratio (around 0.8) and would drop the quality of
plutonium. To counteract this, the reactor is provided with axial and
radial blankets and a spectral shift system using fertile rod clusters.
These features provide a conversion ratio of around 0.9 and also further
reduce the enrichment required. For this purpose, RCVS uses the same
components as current PWR models, apart from the core and associated
equipment (Fig.3.2.2). Fuel rods are standard French PWR Zircaloy cladding
with an outside diameter of 9.5 mm. The lattice pitch is very close to that
of the present 17 x 17 array, but the fuel assembly structure has been
changed to a hexagonal lattice (Fig.3.2.2, 3.2.3). With the plutonium (MOX)
core, the moderator-to-fuel ratio Vm/Vf is about 1.1. In the uranium core,
a number of fissile rods have been replaced by water-filled Zircaloy tubes
having the same outside diameter in order to achieve Vm/Vf of about 2 and
keep the same hydraulic condition.
FERTILE DRIVE HOO
(HYDRAULIC CONTROL MECHANISM)
ABSORBING DRIVE ROD
{ELECTROMECHANICAL
CONTROL MECHANISM)
FERTILE ROD GUIDE
ABSORBING ROD GUIDE
PLENUM SPACER
HEAVY REFLECTOR
FUEL ASSEMBLY EQUIPPED
WITH ABSORBING RODS
WITH FERTILE RODS
FIG.3.2.2. Simplified longitudinal cross-section of the RCVS reactor.
""
GUIDE TUBE
^ OWrr^r^r^^^/^XJoL
URANIUM CORE:WATER ROD
FISSILE ROD
PLUTONIUM CORE:FISSILE ROD
FIG.3.2.3. Transversal cross-section of the RCVS fuel assembly.
Spectral shift rod cluster control assemblies consist of fertile
depleted uranium rods. They are inserted into the core at the beginning of
the cycle in order to harden the energy spectrum of the neutrons. Plutonium
is generated in these rods. As the fuel burns up, rod clusters are withdrawn
in sequence. This system allows Vm/Vf to be varied as follows:
Plutonium Core
Uranium Core
Vm/Vf with
rods inserted
rods removed
Around the core, a heavy stainless steel reflector can be installed to
reduce neutron leakage and limit exposure of the reactor weld area. For the
uranium core, in order to obtain a mean discharge burnup of 45 000 MWd/t, an
initial enrichment of 3.25% is required (4.2% for current PWR). In this
case, there is a gain of 25% with regard to natural uranium consumption and
21% with respect to the fuel cycle cost. The results encourage to envisage
a burnup of 60 000 MWd/t. In this case, the gains in uranium consumption
and fuel cycle cost would amount to 33% and 27% respectively.
For a plutonium core, to achieve a mean discharge burnup of 45 000
MWd/t for mixed plutonium-uranium fuel, an initial enrichment to 5% of
fissile plutonium is required. Under these conditions, the cycle cost for
this type of core is 25% less than for a current PWR. If the discharge
burnup is pushed up to 60 000 MWd/t, enrichment has to be raised to 6%. The
total drop in cycle cost is 30%. In both cases, the plutonium generation
ratio is 0.98. A thermal-hydraulic study of the core shows that the DNB
(Departure from Nucleate Boiling) ratio for this type of reactor differs
little from an ordinary PWR. This ratio can be improved by using mixing
grids. A loss-of-coolant accident study shows no specific difference from
the standard PWR.
3.2.2 Safety Analysis
For RCVS core two particular safety aspects have been concerned.
3.2.2.1 Voidage Coefficient Analysis
The void coefficient (and also moderator temperature effects) can
become positive if the Pu content is too high. This void coefficient in
fact results from the cancellation of large positive and negative capture
and fission cross section contributions. The analysis shows that with an
enrichment lower than 6% of fissile plutonium, the total voidage of the core
would provide overall negative reactivity.
LOCA Analysis
The RCVS design leads to specific problems in case of a loss of coolant
accident (LOCA). The use of a hexagonal matrix, even if its pitch is on the
same order as that of a PWR, increases the pressure loss of the core. It is
thus necessary to demonstrate that safety in case of a LOCA is not
threatened by such a change. The analysis show that the evolution of the
cladding temperature during reflooding is unfavorable for the RCVS. As
compared to a Model N4 NSSS reactor, at the end of decompression the maximum
cladding temperature is about 300°C lower in the RCVS. However, this
advantage is lost during the reflooding phase, and at the end the maximum
cladding temperature attained is on the same order of magnitude.
Research and Development for RCVS
Core Physics
For under-moderated lattices with spectra between those of PWRs and FBR,
most fissions and neutron captures arise in the epithermal range where data
are not well known. For adaptation of computer codes, the major points are
a new version of cross section data library, improvement of resonance self
and mutual shielding calculation and implementation of a specific module for
hexagonal collision probability calculation. Analysis has shown the
importance of the first resonance self shielding for plutonium isotopes 240
and 242. It was also necessary to modify the Pu-242 first resonance
parameters. The computer codes for cell and rectangular or hexagonal
assemblies are under development.
An extensive experimental programme is being carried out in order to
reduce the uncertainties on the different neutronic parameters, which
core parameter studies in tight lattice,
capture cross-section measurements on main heavy nuclides,
total fission product cross-section measurement.
Thermohydraulic Studies
The Thermohydraulic programme has been built in order to fullfil the
lack of data in the three following points.
Boiling crisis in rod bundles with triangular array of different
pitches with and without unheated rods; these coefficients are used
in two-phase flow to compute the local conditions for boiling
crisis data.
Reflood heat transfer and quench front progression in tight lattice
fuel assemblies; preliminary reflooding experiments at imposed
inlet flow rate have shown that tight lattice cores are more
difficult to be cooled than standard PWR ones. An effort would
then be useful which could consist in improving the system effect
so as to get more liquid flow rate entering the core or in
increasing core heat transfer during the refilling phase.
Coolability of fertile rods inserted in guide tubes by water
circulating in a narrow gap; up to now the tests show that the
boiling crisis is not dangerous for the cladding of the fertile rod.
These programmes have been performed in the Thermalhydraulic Laboratory
of the Nuclear Research Center of Grenoble.
THE CONVOY PLANTS (PWR 1300) [48,49] (FEDERAL REPUBLIC OF GERMANY)
The Convoy concept
The Convoy plants are a group of three plants with PWR of the standard
size for Germany of 1300 MWe (net), in the Federal Republic of Germany,
which are presently under construction almost in parallel at three sites in
the FRG, for as many utilities. These nuclear power plants are a continuous
development from the precursor projects of the same unit size, such as
Philippsburg 2, Grohnde and Brokdorf, the former 2 being commissioned in
1984 and the latter in 1986. The Convoy concept was established in 1980.
The advanced features of the Convoy concept lie in the field of the
engineering and project management associated with nuclear power plant
construction. The concept features:
detailed preliminary planning prior to commencement of construction,
reduction of the engineering effort per plant,
streamlining of specifications and procedures,
sharing of tasks between Authorized Inspection Agencies,
economical manufacture of large numbers of identical components,
rationalization of licensing procedure.
The concept also included reorganization of the specifications to
differentiate between different requirement categories, to adapt quality
control measures, to suit the manufacturing process and to reduce the amount
of documentation. Three partial construction permits and one operating
licence were envisaged for the licensing procedure. The first partial
construction permit was to cover the concept and the civil engineering part,
the second the entire mechanical and electrical part and the third the
initial loading of the core.
For a series of successive nuclear power plants, the procedure provided
for uniform planning using identical software and hardware for all
site-independent areas of the plants. In the process, the planning work was
to be done sufficiently early to ensure that the partial construction permit
for all the mechanical and electrical systems could be issued before the
commencement of erection work. This approach made it possible to reduce the
amount of engineering manhours required and to stabilize prices by placing
large orders for components of identical design. Fig.3.3.1 shows the degree
of standardization of the Convoy plants, which are subdivided into a
standard and a site-specific part. Most of the site-specific facilities
concerns the cooling water systems and the connection with the power grid,
where the differences are unavoidable.
Site-independent
Emergency power generating building UBP
Swltchgear building UBA
• Emergency feedwater
bullding ULB
r Owner's facilities
Plot plan, main plant
Circulating water pump
building, Cooling tower
Reactor auxiliary
Site-dependent
pojjj'
FIG.3.3.1. Degree of standardization of Convoy plants.
Sharing of tasks between Authorized Inspection Agencies means that a
specific aspect of the identical design for the main plant buildings and
systems was reviewed and approved by one of the Inspection Agencies
participating in the Convoy licensing procedure, and this approval was
accepted by the other Inspection Agencies. The special conditions,
stipulated in the construction permits for the mechanical, electrical and
instrumentation and control systems, are reduced to less than 10% of those
for earlier plants. Those conditions imposed usually require changes to the
original plans and result in additional costs or delays in construction.
The reduction of such conditions is an impressive demonstration of the
success of the Convoy concept.
Planning for pipe routing in the reactor building was not only
considerably shorter for Convoy plants than for earlier plants, but was also
practically completed when construction began. The very comprehensive and
detailed planning was accompanied by continuous quality control. For
example, the arrangement of components, pipework, cables and ventilation
ducts in the buildings were planned using models on a scale of 1 : 25.
Finishing work such as framework removal, laying of floor topping, which
were previously planned as a whole, were now broken down into individual
activities with detailed time scheduling and performed immediately after
completion of the concrete. The préfabrication of piping is increased, for
example, in the reactor coolant lines, 9 of the 15 welds per loop are shop
welds. An additional aid was the survey data for all the anchor plates and
their actual points of attachment to the building structure in the main
buildings prior to commencement of the actual erection activities. In this
way the construction time for a 1300 MW PWR Convoy plant could be reduced to
approx. 60 months.
Technical development of Convoy plants
The technical development of Convoy plants focused on design details
and optimization of plant and building layout. The reactor building annulus
(space between the spherical steel containment shell and the outer concrete
structure) was enlarged by 4 m in diameter in order to rearrange the
mechanical equipments and obtain physical separation of redundant components
and systems, fire protection, better accessibility and maintainability. The
reactor auxiliary building interior was redesigned to improve accessibility
for inservice inspections and maintenance etc. In the turbine building the
arrangement of the filters and piping have been optimized to reduce
corrosion-product ingress into the steam generators.
For reactor pressure vessel the number of circumferential welds could
be reduced from eight to five and longitudinal welds are abandoned. A 100%
inservice inspection of RPV could be performed within five days. The reactor
core internals are welded compared with the previous design in which the
different parts are connected by screws. The design improvements of steam
generator led to increase in the weight of the SGs by about 20% from
previous plants to Convoy series with a higher design pressure on the
secondary side. The reactor coolant pumps use forge casings. The sealing
system consists of three independent seals, one of them is a stand still
seal which prevents small leaks even in the event of failure of the seal
water supply. The reactor coolant lines were completely prefabricated, only
the connecting welds to the primary components have to be performed on
site. The introduction of the leak before break criterion allowed to
eliminate the need for pipe whip restraints. The steel containment is
constructed from material 15 MnNi63 with improved weldability.
containment is designed to contain the maximum pressure which can arise
under accident conditions. The containment is divided into the operating
compartments which are accessible during reactor operation, and the plant
compartments which are not accessible during reactor operation. It also
contains the spent fuel pool. The polar crane manipulates the spent fuel
transfer casks in principle allow the replacement of major components,
including steam generators.
Emergency core cooling and residual heat removal systems have 4 trains
redundancy. Each loop is equipped with two accumulators, one of which feeds
into the hot and one into the cold leg. Each loop has its own borated water
storage tank. In addition to the main feedwater pumps there are total of 6
pumps available for delivering feedwater, which can be driven by the station
service power supply and by the diesels. Therefore the feedwater supply has
an extremely high reliability.
The special features of instrumentation and control systems are
30-minute-criterion and limitation system. The 30-minute-criterion calls
for all actions necessary after the reactor protection system has responded
to run automatically for the first 30 minutes. A limitation system
intervenes before the reactor protection system response in order to return
the reactor to normal operation condition. It is only if this limitation
system fails to prevent the reactor that the reactor protection system takes
over. A significant improvement in the supply of information to the
operators is the process information system PRINS which went into operation
for the first time in the Convoy plants. This system makes large-scale use
of full-graphic-capability VDUs (video display units) which can be termed an
"expert system" incorporating a data base (signals, computer results), a
knowledge base (information goals, plant and computing functions) and an
inference "engine" which reasons with this knowledge. It allows both
optimization of operation and the detection at an early stage of small
deviations.
3.3.3 Safety Aspects
For Convoy plants the safety goals and requirements are identical to
those of the pre-Convoy plants. However, experience from TM1 and the German
Risk Study was utilized to perform specific improvements in a number of
areas. These consisted in particular of measurement of the coolant fill
level in the RPV, the display in the control room of at-a-glance information
on the subcooling margin in the reactor coolant line, and installation of
systems for limiting the hydrogen concentration in the containment after a
loss-of-coolant accident.
The leak-before-break criterion was introduced for the piping of the
pressure retaining boundary and of the main steam and feedwater lines.
Introduction of the "Basic Safety Concept" philosophy has been a major
influence of the compilation of the specifications valid for today's Convoy
plants. Under the heading "basic safety", the components of the
safety-related systems were subjected to a series of improvements in terms
of mechanical design, material selection, stress limitation, quality
assurance and ease of inservice inspection. In order to protect the
containment integrity in case of core melt accident a pressure relief
facility with filter system was installed.
THE SIEMENS 1000 MWe THREE LOOP PWR [50-52] (FEDERAL REPUBLIC OF GERMANY)
The Siemens 1000 MWe Loop PWR design was derived from the four-loop
plant operating in the Federal Republic of Germany. The plant power range
chosen is the most common one. The new plant follows closely
internationally applied safety and licensing practices, while conforming to
the basic German safety regulations and design criteria. The plant design
suits a number of international proposed sites or can easily be adjusted to
the requirements of the particular local conditions. The main parameters of
the plant are listed in Table 3.4.1.
3.4.1 The Core Design
The reactor core is made of 177 geometrically identical fuel assemblies
in a 18 x 18 - 24 square configuration. In the first core part of the
excess reactivity is compensated by Gd2U3 as burnable absorbers. The
Gd203 fuel rods contain natural uranium as a carrier over their entire
active height. A flat power density distribution throughout the first core
is attained by using fuel assemblies with three different enrichments. The
axial power density distribution is flattened by the use of burnable
absorbers which do not extend to the upper and lower ends of the active
core. A flat power density distribution can be maintained not only at
constant-load operation, but also during load-change operation.
For fuel management, an in-out strategy is preferred. In this case,
the use of gadolinium oxide is particularly advantageous because it burns
out completely during its first residency period. Separate waste management
for the absorber rods is avoided as Gd203 is homogeneously mixed with
the fuel of several fuel rods. The improvement in neutron economy brought
about by in-out fuel management increases the equilibrium cycle length by up
to 20 full power days, the savings in reload enrichment amounts up to 0.15
wt% U235.
The power density distribution is monitored by the incore
instrumentation. Two independent systems, the aeroball system and the
fixed-position self-powered detectors, are provided. They support and
complement each other. A process computer prints a complete
THE MAIN PARAMETERS OF THE PLANT
Thermal reactor output
Thermal steam generator output
Net electrical output
Primary Containment Vessel
Design overpressure/temperature
4.9 bar/145°C
Reactor Pressure Vessel
- Inside diameter cylindrical shell
- Wall thickness of the cylindrical shell
- Design pressure/temperature
- Weight without internals
bar/350oc
Reactor Coolant System
Number of assemblies
Fuel rods per fuel assembly
Overall length of fuel rods
Active length of fuel rods
Outside diameter of fuel rods
Overall in-core uranium weight
Number of coolant loops
Reactor operating pressure
Coolant inlet temperature
Coolant outlet temperature
Coolant flow rate
Sintered UO2
82.3 t
293.8°C
15 876 kg/s
Tube material
Design pressure/temperature
Overall weight
21 500 ram
87.3 bar/350°C
Reactor Coolant Pumps
Design flow rate
Motor rating cold/hot
5292 kg/s
9870/7320 kW
Pressurizer
Diameter (inner)
- Volume
13 800 mm
three- dimensional image of the power density distribution from approx. 800
measured values within 10 minutes of completion of aeroball measurement.
There is on-line continuous DNB surveillance with the core protection system
(CPS). The CPS safeguards a DNBR dependent power limit. The main advantage
gained from the use of the core protection system is that both core
surveillance and fuel management apply the same design parameter: DNBR.
The core protection system and fuel management together meet the stringent
requirements associated with the DNBR safety limit, yielding a sufficient
increase in the DNB margin to permit implementation of full low leakage fuel
With fuel assemblies of the type 18 x 18 - 24 a high maximum local
burnup is attainable. As a result, long operation cycles up to 18 months
and high reload burnups can be achieved, thus minimizing the number of
reload fuel assemblies and, as a consequence, also the power generation
costs.
After excess reactivity is exhausted, the reactor can either be
shut down for refueling, or it can run continually at reduced power in the
stretch-out operation mode.
The potentially most important technological limitation on design
burnup in modern PWRs with high thermal hydraulic efficiency would appear to
be waterside Zircaloy corrosion. With this in mind, a model for corewide
analysis of waterside corrosion was developed by Siemens which takes into
account plant parameters as well as individual fuel rod power histories.
The very sophisticated 3-dimensional code system permits local oxide layer
thickness to be evaluated for each fuel rod and then utilized as a special
optimization parameter.
3.4.2 Load Follow Capability
The Siemens PWR plants have demonstrated their load following
capability and the possibility for frequency control. The unrestricted load
follow capability constituted a fundamental design criterion for German NPPs
from a very early stage in the development of the nuclear programme, this
applies to all parts of the nuclear steam supply system, particularly for
primary and secondary system engineering, inclusive of the turbine
generator unit,
instrumentation and control systems,
mechanical components,
nuclear auxiliary systems.
Attention was also paid to core monitoring systems and to reactor
controls, as these provide a major contribution both to load follow
capability and to economic fuel utilization. The PWR plants feature an
Integrated Power Control and Reactor (Core) Limitation and Protection System
(Fig.3.4.1). The load follow operation can be performed in an entirely
automatic manner from the control room where the operator simply selects the
desired ramp and the power level to be reached. Integrated monitoring and
control systems assure that the plant remains within its technical
specifications at all times. For the case of a loss of generator load
followed by a turbine trip the plant can be brought again to full power in
about 30 minutes. A further ability is to change over to part load
operation instead of plant trip in the event of faults within the power
plant (e.g. main coolant pump trip).
For the reactor controls, the control assemblies (CA) and boron
poisoning in the coolant form the final control elements. The Siemens
control assemblies (CA) management scheme in which there is only one type of
Deminemliieal
Bent acid
Closed control loops lor
L bank position
D bank position
Control assembly actuation
Generator active power
Valve position
Minimum main steam pressure
Maximum main steam pressure
0 bank position
L control assembly bank
D control assembly bank
FIG.3.4.1. Block diagram of the power control systems of a KWU nuclear power plant with pressurized
water reactor.
CA, is much less complicated than the other type of concept (Fig.3.4.2). In
functional terms, the CAs are grouped into two banks. The weak (reactivity
worth) D-bank is preferably used to control integral reactor power and the
strong L-bank is used to control the axial power distribution. The L and
the D-bank move in opposition to each other, such that their effects on
total power cancel out. Operating experience has shown that the Siemens
control concept does not require the use of grey CAs.
L-bank
|| || ||
D-bank *)
Shut down bank
Regulating bank *)
Grey CAs
Part Length
rods (sometimes)
*) D-bank and regulating bank consist of a number of sub-banks
which are inserted in sequence with decreasing power
= Group of CAs moving together to perform specific tasks
FIG.3.4.2. Control assembly (CA) configurations for upper power range.
In order to handle the transient condition during the load following
operation, the Siemens PWR plant has a unique feature of a three-stage core
control and protection system. Any parameter subject to design or safety
limits, such as peak local power density, is normally kept within a certain
"control band" by the automatic control system. If the limits are exceeded,
redundant core protection systems (known as "limitation systems"
automatically bring the parameter back into the control band without
interrupting operation. Only if the control and the limitation systems fail
to handle the transient condition does the automatic reactor trip system
come into play. This kind of control and limitation with graded
counter-measures is applied on both global core parameters (nuclear and
thermal-hydraulic limits) and local power densities in the upper and lower
part of the core (limits on peak power density, departure from nucleate
boiling (DNB) and pellet clad interaction (PCI)). The local core protection
and control systems use signals from in-core detectors. The global core
protection functions mainly use signals from the out-of-core instrumentation.
For new plants, an additional redundant automatic limitation function,
which protects the fuel against unacceptable local PCI loads during severe
transients, is equipped. The maximum allowable power density with respect
to PCI is calculated by adding the actual preconditioned power and a given
value for the overshoot allowance. The result is a "sliding PCI limit".
The sliding PCI limit is an integral part of the automatic power density
limitation system and is set in parallel with other limits, such as DNB and
LOCA limits. The great advantages of this system are that not only is the
core protected against PCI risks, but also the operator is completely
relieved from PCl-related core surveillance tasks.
Improvements of Systems and Components Design
The improvements of system and component design are aimed at increasing
the plant reliability and availability bases on all the experience gained
from the operating plants. The number of valves and the length of tubes
have been optimized in the sense of easy operability and maintainability and
economic viability of the entire plant. These factors also have been
considered for the design of the instrumentation and control system and the
number of electric drives. The main systems and components have been left
unchanged.
The reactor building is constituted as double containment. The reactor
vessel is made of forged rings to eliminate axial welds at the reactor
beltline region where radiation fluence is high, and to minimize in-service
inspection.
3.4.4 Safety Systems
In accordance with the requirements of the Guidelines of the German
Reactor Safety Commission for Engineered Safety Systems, the following
design principles are applied:
redundancy, diversity, general avoidance of interconnected systems,
physical separation of redundant trains,
fail-safe operation of systems during failure of subsystems and
plant components.
An (n + 2) redundancy is adopted, where n is the number of engineered
safety trains. This (n + 2) redundancy ensures that even in the event of a
single failure and one train being out of operation for maintenance, the
full capacity of the safety system concerned is available.
By analogy with the 1300 MWe Standard PWR (four-train safety system,
one train being connected to each of the four loops), the 1000 MWe
three loop design is provided with three-train engineered safety systems
which are connected to each loop without interconnections. This three-train
safety system design also complies with the (n + 2) redundancy requirement.
Protection against failures caused by events such as fire or flood is
achieved by the redundant trains of a safety system being physically
separated from each other or structurally protected. The redundant and
physically separated arrangement of the trains is also applied to their
emergency power supply, the necessary auxiliary systems and the actuation of
their functions by the safety-related instrumentation and control. This
ensures a high reliability of the engineered safety systems.
Pipe break philosophy
Reactor Safety Commission, quality assurance measures are taken to ensure
that only subcritical leaks can occur in reactor coolant lines (maximum leak
cross section equivalent to 0.1 A). Examples of such measures are as
use of high-quality materials, in particular with respect to
ductility,
conservative limitation of stresses,
prevention of stress peaks by way of optimized design and
assurance of the application of optimized manufacturing and testing
knowledge and assessment of faulted conditions,
consideration of the coolant quality.
This leak postulate forms the design basis with respect to the load
assumptions for reaction and jet forces on pipes, components, component
internals and buildings.
A leak cross section corresponding to an area equivalent to a
double-ended pipe break (2 A) is postulated for the design of the emergency
core cooling system, for determination of the containment design pressure as
well as of pressure gradients inside the containment.
HIGH CONVERTOR REACTOR (HCR)
[53, 54] (FEDERAL REPUBLIC OF GERMANY)
Siemens has for many years pursued the following important objectives
for improving fuel element and core design as well as fuel management
procedures :
reduction of fuel cycle cost by increasing the average fuel
discharge burnup up to 50 MWd/kg, improving fuel utilization using
advanced fuel element design and fuel management strategies, e.g.,
applying gadolinium burnable absorber and all-zirconium fuel
elements in low-leakage fuel management procedures,
enhancement of operational flexibility by designing the core for a
flexible fuel cycle length up to 2 yr, stretch-out capability, and
load-following capability,
improvement of fuel utilization by recycling reprocessed plutonium
and residual uranium, 20 000 plutonium rods having been irradiated
to local burnups exceeding 50 000 MWd/t.
Various measures have been proposed to improve fuel utilization in
present PWRs having a standard fuel rod lattice and being operated in the
once through cycle mode (APWR). The most effective ones for an interim step
reflector improvements,
use of hydraulically driven spectral shift displacement rods,
application of low-leakage management strategies in combination
with the use of burnable absorber.
However, all these suggestions, even if applied in combination, will
not bring about an ore savings effect exceeding approximately 20%.
Therefore, a logical and natural development of the standard PWR for
achieving a really substantial ore utilization is a Light Water High
Conversion Reactor (LWHCR). The essential advantage of a high converting
reactor is the first one, namely to use as much as possible the basic and
proven design and construction principles and operation experiences gained
over several decades. All the components except core, core internals and
closure head will be the same for the LWHCR and the PWR. Timely commercial
introduction of a HCR would be decisively facilitated if a standard PWR
could be converted into a HCR.
If the above mentioned constraints can be met, it is virtually assured
that capital cost can be kept in the range 1-2% in excess of that for a
conventional PWR. This is a necessary requirement for the commercial
viability of the HCR.
The design objectives for HCR are:
thermal power (HCR) = thermal power (PWR),
conversion rate greater than or approximately 0.9,
discharge burnup up to ~ 70 MWd/kg for the long-range target,
using stainless steel fuel rod cladding,
void reactivity feedback as to meet all licensing rules of the
German reactor safety commission,
coolability in the operation mode and in emergency cases,
reduce the fuel cycle cost approximately 10%.
Table 3.5.1 shows preliminary core design data and Fig.3.5.1 shows the
reactor core and control assembly in comparison with those of a standard
1300 MWe PWR.
HCR CORE DESIGN DATA
Thermal power, MW
No. of fuel assemblies
Pin diameter, mm
Fuel assembly shape
Type of spacer
Active height (m)
Average linear heat rating (W/cm)
Power density (kW/L)
Average water-fuel volume ratio
Average reload enrichment, wt%
helical fins
~ 7.5 +0.2
> or ~ 3 .3
Limited modifications
vessel internals
pressure vessel lid
pump impeller
KHCR
193 Fuel assemblies
61 Control assembly positions
151 Control assembly positions
61 Drive mechanisms
additional Investment cost for
adaptation of a standard PWR to a KHCR
1 - 2 % offne total investment of a PWR
Control assembly position
Reflector elements
'—•> Tl«
, Control assembly
d - 9.5 mm
p = 10.67
Water-to-fuel volume ratio: * 0.5
FIG.3.5.1. Reactor core and control assembly.
Since the water-to-fuel volume ratio is decreased from 2.06 to
approximately 0.50, a fissile enrichment of roughly 7-8% Pufiss and/or
U-235 leading to a fissile inventory of nearly 9 metric tons for a 1300 MWe
PWR plant is required. A homogeneous fuel assembly design is currently
preferred over a heterogeneous seed and blanket concept due to the higher
degree of affinity to the proven fuel technology of the standard PWR. It is
well understood that the degree of safety, reliability and economic
performance achieved by current LWRs is a standard that has to be maintained
or even exceeded with any new reactor type: Safety relates to both
temperature and void reactivity effects and to heat removal during normal
operation and anticipated operational occurances, as well as in postulated
loss-of-coolant accidents.
Although the HCR can be based to the highest possible extent on the
well established standardized PWR component and plant system technology,
certain areas require both analytical and experimental investigation and
verification. These areas relate especially to thermohydraulic, mechanical
and neutron physics core design and associated safety and licensing items,
as well as fuel irradiation performance. The main development items to be
dealt with are:
critical heat flux (DNB) and pressure drop tests,
void reactivity experiments in a zero power critical facility in
order to demonstrate an inherently negative void reactivity
emergency core cooling experiments in order to give evidence of
coolability after a LOCA,
furthermore, questions like the following have to be addressed:
design specifications for size, number and positioning of fuel
and control assemblies,
thermohydraulic and mechanical reflector design,
fuel assembly hold-up device optimization.
It may be expected that a conversion ratio of 0.95 can be achieved with
a negative void coefficient. For thermohydraulics there are certain
indications that cladding tubes with helical fins of optimized inclination
may crucially increase heat transfer conditions to such levels that even in
tight lattices heat transfer can be ensured.
Early stainless steel clad LOCA ballooning experiments performed at the
KfK REBEKA facility showed satisfactory deformation behaviour at high
temperatures. The results have demonstrated a well coolable lattice
geometry after a postulated LOCA and recent flooding experiments performed
for a very tight lattice with p/d = 1.06 at the FLORESTAN facility (KfK)
suggest that even extremely tight rod configurations appear principally
coolable. Results from additional tests at the NEPTUN (EIR) facilities
based on rod configurations with p/d approximately 1.12 will be the basis
for computer code adaptation and verification within the development
cooperation. Various SS single test rods with and without fuel are
scheduled for irradiation at the Obrigheim PWR station to check the
suitability of the envisaged clad material for high burnups. Test fuel
bundles and fuel assemblies are scheduled for irradiation in a power reactor
in the near future. Mechanical design problems will have to be solved for
all components within the pressure vessel. Design details of all these
components influence the integral concept. The overall objective of this
R&D programme is to have the technical feasibility, including that for
licensibility, established by the early 1990s as a prerequisite for the
decision whether to enter a demonstration plant programme.
ADVANCED BWR 90 [55] (SWEDEN)
BWR 90 is the ABB-ATOM 1000 MWe nuclear power plant for the 1990s,
based on the design and operation of BWR 75 in Finland and Sweden. The BWR
75 design developed in the 1970s is characterized by the use of internal
recirculation pumps, fine motion control rods, and extensive redundancy and
separation of safety-related systems. The experience of these plants forms
the basis for the design of the BWR 90. Moderate modifications have been
made to adapt to updating technologies, new safety requirements and to
achieve cost savings. The average capacity factor of BWR 75 plants is about
90%, the occupational radiation exposure is about 0.5 manSv (1986 figures).
Reactor Design
Traditional recirculation system and favourable load-following
The reactor design has not been changed much. The recirculation system
is based on the internal pumps driven by wet motors of the glandless
squirrel cage asynchronous type. The motors are supplied individually with
"variable frequency-variable voltage" power from frequency converters. This
type of pump has been operating reliably (for more than two million
operating hours) since 1978.
The internal pumps provide means for rapid and accurate power control
in the high power range, and they are also advantageous for load-following
purposes. The plant is characterized by the capability to accept a 10% step
change in power with an equivalent time constant of 15 s with the reactor at
constant pressure, or 5 s with floating pressure control. Ramp load changes
of 20% per minute is accepted and useful for all operating plants of BWR
75 model. In the high power range, between 70% and 100% of nominal power,
daily variations can be accommodated with the change rate above without
restrictions. For wider power variations the extended range is achieved by
means of adjusting the control rod pattern. Daily load following down to
40% is easily accommodated this way with a power reduction ramp of one hour
or less. The return to full power from 407o will take two hours, taking
current operating restrictions into consideration, but this is usually quite
acceptable for the grid requirements.
Weekend load following, for example to meet reduced demands during
weekends, may require a more cautious return from 85% to full power
depending upon the past history of the movement of the control rods and the
preconditioning of the fuel. The reason is that for the longer periods at
reduced power, the xenon content in the fuel will reach a lower level and
the return to full power will mean restoration of the Xenon content.
Depending upon the details of the control rod sequence, it may be possible
to reduce the waiting periods at 85% and 95% of full power. Current
development work on nuclear fuel will most probably soon make it possible to
ease the operating restrictions considerably.
The internal recirculation pumps have more than 10% excess flow rate
capacity, which allows xenon override, and the fine motion control rod
drives and the grey-tipped control blades allow control rod movements at
full power. The built-in redundancy in the internal recirculation pump
system implies that the reactor can be operated at full power even if one
pump should fail. These load follow characteristics and the capability of
operation at full power with one recirculation pump out of operation have
been successfully demonstrated in the operating plants.
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