HPLWR – D 13
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EUROPEAN COMMISSION
5th EURATOM FRAMEWORK PROGRAMME 1998-2002
KEY ACTION : NUCLEAR FISSION
n
HIGH PERFORMANCE LIGHT WATER REACTOR (HPLWR)
CONTRACT N° FIKI-CT-2000-00033
SUMMARY REPORT OF THE HPLWR PROJECT
(HPLWR Deliverable D 13)
D. Squarer (FZK, Karlsruhe, Germany), D. Bittermann (Framatome ANP,
Erlangen, Germany), Y. Oka (U. of Tokyo, Tokyo, Japan), P. Dumaz (CEA,
Cadarache, France), G. Rimpault (CEA, Cadarache, France), R. KyrkiRajamaki (VTT, Espoo, Finland), K. Ehrlich (FZK-MCS, Karslruhe,
Germany), N. Aksan (PSI, Würelingen, Switzerland), C. Maraczy (KFKI,
Budapest, Hungary), A. Souyri (EdF, Chatou, France)
Dissemination level :
RE: restricted to a group specified by the partners of the HPLWR project
October, 2002
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TABLE OF CONTENTS
Summary ........................................................................................................ 4
1. Work Package I – Evaluation of current status and definition of
basic requirements ........................................................................................ 7
1.1
Objectives .................................................................................................................................... 7
1.2
Description of Work .................................................................................................................... 7
1.3
Deliverables and milestones ........................................................................................................ 7
1.4
Methodology ............................................................................................................................... 7
1.5
Review of other concepts ............................................................................................................ 8
1.6
Design requirements .................................................................................................................... 9
1.7
Proposed design ........................................................................................................................... 9
1.7.1
Primary System......................................................................................................................10
1.7.2
Containment and Safety concept ...........................................................................................11
1.7.3
Balance of plant (BOP) ..........................................................................................................12
Table 1.1- Proposed Characteristics of the HPLWR Power Plant .............................................................14
1.8
References ..................................................................................................................................15
2. Work Package II- Core design and theoretical analyses .................. 20
2.1
2.2
2.3
2.4
2.5
2.5.1
2.5.2
2.5.3
2.6
2.6.1
2.6.2
2.6.3
2.6.4
2.7
2.8
Objectives ...................................................................................................................................20
Description of Work ...................................................................................................................20
Deliverables and Milestones .......................................................................................................20
Methodology ..............................................................................................................................20
Results of benchmark problem ...................................................................................................21
2-D subassembly calculations ................................................................................................21
Core calculations ...................................................................................................................23
Conclusions on computer codes for HPLWR ........................................................................24
Shortcomings And Proposed Modifications To The Fuel Assembly .........................................25
Design criteria........................................................................................................................26
Cladding materials .................................................................................................................26
Fuel assembly proposals ........................................................................................................27
Solid moderator .....................................................................................................................27
Conclusions.................................................................................................................................28
References ..................................................................................................................................28
3. Work Package III – Reactor Safety and Deteriorated Heat Flux .... 30
3.1
Objectives ...................................................................................................................................30
3.2
Description of Work ...................................................................................................................30
3.3
Deliverables and milestones .......................................................................................................30
3.4
Assessment of Required Safety Features ....................................................................................30
3.4.1
Containment Concept and passive safety features .................................................................31
3.4.2
Safety Concept .......................................................................................................................31
Table 3.1 – Proposed passive and active safety systems for the HPLWR .................................................32
3.5
General Application of Some Safety Requirements ...................................................................33
3.5.1
European Utility Requirements for Safety versus HPLWR ...................................................34
3.5.2
Generation IV Technology Goals in The Safety and Reliability Area versus HPLWR ........34
3.6
Deteriorated Heat Transfer in Supercritical Water .....................................................................34
3.7
Proposed Design .........................................................................................................................35
3.8
Preliminary Transient Safety Analyses of the HPLWR .............................................................35
3.8.1
Use of RELAP5/Mod3 Computer Code ................................................................................35
3.8.2
Use of CATHARE 2 Computer Code ....................................................................................37
3.8.3
Use of TRAB Computer Code ...............................................................................................38
3.8.4
Future work ...........................................................................................................................38
3.9
Conclusions ................................................................................................................................39
3.10
References ..................................................................................................................................39
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4. Work Package IV-Summary Report on Material Selection and
available Treatments for Reduction of Corrosion in HPLWR
components. ................................................................................................. 41
4.1
Objectives ...................................................................................................................................41
4.2
Description of work ....................................................................................................................41
4.3
Deliverables and Milestones .......................................................................................................41
Table 4-1:
HPLWR ”Reference Design” Data for in-core, RPV, and
ex-core components ........45
Table 4-2:
Estimated maximum temperatures for different materials for the condition of R M/45,000 h at
100 MPa and 200 MPa respectively ..........................................................................................................46
4.4
References ..................................................................................................................................49
5. Work Package V – Economics ............................................................. 50
5.1
Objectives ...................................................................................................................................50
5.2
Description of Work ...................................................................................................................50
5.3
Deliverables and Milestones .......................................................................................................50
5.4
Review of other economic studies ..............................................................................................50
5.4.1
SCLWR economic study .......................................................................................................50
5.4.2
Comparison with other existing concepts ..............................................................................51
5.4.3
GE’s Advanced BWR for improved economics ....................................................................51
5.4.4
DOE’s Near Term Deployment Economics ..........................................................................52
Table 5.1 Cost comparison of ALWR and gas turbine plants ...................................................................52
5.5
Considerations concerning the economics of a HPLWR plant ..................................................54
5.5.1 General ........................................................................................................................................54
5.5.2
Fuel Cycle Considerations .....................................................................................................55
5.5.3
Specific evaluation of HPLWR electricity generation costs ..................................................55
5.4
Conclusions ................................................................................................................................56
5.5
References ................................................................................................................................57
6. Conclusions ............................................................................................ 58
Appendix A : List of HPLWR Reports, Minutes and Memos................ 60
LIST OF FIGURES
Figure 1. 1 -Example of Hexagonal Fuel Assemblies ..................................................................................16
Figure 1. 2- Reactor Pressure Vessel (RPV) and Arrangement For Inlet and Outlet Nozzles .......................17
Figure 1. 3- HPLWR Containment for a 1000 MWe Plant ...........................................................................18
Figure 1. 4- Schematic of the HPLWR Circuit Diagram ...............................................................................19
Figure 3. 1– Containment and Primary Circuit Concept for the HPLWR .....................................................33
Figure 4- 1
A comparison of oxidation and spallation for ferritic and austenitic steels at 600°C. Metal
loss is half of the oxide thickness. .........................................................................................................47
Figure 4- 2
Ultimate tensile strength RM for selected alloys as a function of temperature .......................47
Figure 4- 3
Creep-rupture strength RM/45,000h for selected alloys ..............................................................48
Figure 4- 4
The evolution of the mean hoop stress in the cladding by comparing the effect of outer
corrosion with the combined outer and inner corrosion ........................................................................48
Figure 5. 1
Cost comparison of Advanced LWR (ALWR) with Combined Cycle Gas Turbine (CCGT)
and Gas Turbine (GT) (DOE Near-Term Deployment, October 2001) .................................................53
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Summary
D. Squarer(FZK)
The HPLWR project objectives are: (a) to determine the state of the art of the technology
with relevance to the HPLWR conditions, (b) to determine the technical merit and
economic feasibility of an HPLWR, (c) to identify the main difficulties that may lie
ahead, (d) to recommend future R&D program if the concept is found to be feasible.
The project was organized into six Work Packages (WP): WP I- Plant definition and
architecture, WP II- Core design and theoretical analyses, WP III - Reactor safety and
deteriorated heat flux, WP IV – Materials and corrosion, WP V- Economics, WP VIProject management. Communication and exchange of information between the WPs was
achieved through five general project meetings in which the results of every WP were
discussed. Additional meetings were held by specific WPs. Detailed Minutes of the
project meetings were prepared in order to help disseminate the necessary information.
Substantial amount of technical information was generated and documented by the
HPLWR project, as demonstrated by the reference list of Appendix A. A brief summary
of the major results of the project is given in this report by each WP.
The following accomplishments can be highlighted at the conclusion of the HPLWR
project :
 A review and assessment of the state-of-the-art of supercritical-water cooled
reactors, as well as relevant technolgical review of supercritical fossil power
plants, has been performed and its results were considered during the execution of
the HPLWR project. These results indicate that the once-through reactor concept
outlined by Prof. Oka of the University of Tokyo, could prove to be economically
and technically competitive with other advanced LWRs as well as with fossil
power plants. Consequently, this concept, which contains similarities to existing
and advanced LWRs design in Europe and Japan, was selected by the HPLWR
project as a “reference design” in order to assess the technology and the available
tools for the analyses of supercritical-water cooled reactors.
 General plant characteristics of a 1000 MWe once-through supercritical water
reactor power plant, that has a potential to be economically competitive, were
defined in WP I (Table 1.1). Preliminary concepts for a fuel assembly, pressure
vessel, containment and circuit diagram were also defined (Figures 1.1-1.4).
 Extensive neutronics and thermal-hydraulics core calculations were carried out in
WP II on a benchmark problem generated from the “reference design” and on
potential fuel assemblies for the HPLWR. Independent calculations were carried
out by several partners with different codes in order to: verify the analyses,
identify computer codes that could analyze the HPLWR core, identify any
required code development effort and identify shortcomings in the design itself.
Several shortcomings were identified in the fuel assembly of the “reference
design” (e.g. under-moderation, excessive neutron capturing by structural
material, short burn-up, etc.) and these findings could guide the design of an
improved fuel assembly. Additional effort has to be invested in order to complete
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the development and verification of the various computer codes, to design an
improved fuel assembly and to complete a whole core analysis with consistent
assumptions.
The general safety features and safety philosophy of the HPLWR were defined in
WP III and computer codes that could perform the safety analyses of the HPLWR
were identified and tried under supercritical water conditions. The safety
philosophy is based on existing and advanced LWR designs and it follows the
European Utility Requirements (EUR) and Generation IV guidelines and criteria.
Although the information described in WP III is general and very preliminary, it
supports the contention that the HPLWR may be designed to operate safely and is
expected to reach the safety level of advanced LWRs. At the conclusion of the
HPLWR project, the design has not been completed in sufficient details to allow
accurate safety analysis, the expected regulations have not been explored and the
computer codes that could be used to perform safety analysis have not been
validated and verified under supercritical water conditions. Only after the
completion of these tasks, can an accurate safety analysis of the HPLWR be
completed. Nevertheless, very simple and preliminary results obtained by these
safety analysis codes, indicate that they could support the introduction and design
of appropriate safety systems, and that they would be able to perform accurate
safety analysis after additional code development. In support of the fuel assembly
design, a thorough review of heat transfer at supercritical pressures was
completed together with a thermal-hydraulics analysis of potential HPLWR subchannels. These results will be used in the design of improved fuel assemblies.
In WP IV a state-of-the-art study was performed to investigate the operational
conditions for in-vessel and ex-vessel materials in a HPLWR and to evaluate the
potential of existing structural materials for application in fuel elements, core
structures, reactor pressure vessel and out-of-core components. Based on
extensive past experience of material behavior in LWRs, fast breeder reactors,
supercritical fossil power plants, and supercritical waste oxidation, the partners
were able to recommend in WP IV promising HPLWR materials for in-vessel (up
to 650 ºC) and ex-vessel applications that could be strong enough at the design
temperature and also possess reasonable corrosion resistance characteristics. The
in-vessel material selection was done in close cooperation with WP II in order to
identify potential materials that are neutronically compatible. The preliminary
identification of potential materials (Table 4.2) must be verified by additional
analyses and extensive testing (in particular for corrosion) since the applicable
data base is totally inadequate.
The economic evaluation in WP V was performed by first reviewing the
economic study of the “reference design” that was performed for Japanese
utilities by comparing the cost of the “reference design” with the cost of the
ABWR. Furthermore, measures that were taken by the industry to improve the
economics of Advanced BWR were reviewed, and the cost of generating
electricity by fossil power plants was highlighted as the competitive cost “to
beat”. In addition to the HPLWR being a more compact (smaller RPV and
containment) and simpler, several components used by LWRs are not required by
the HPLWR (e.g. steam generators, recirculation pumps, steam separators,
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pressurizer). A significant economical benefit can be obtained when ‘off-theshelve’ equipment is used. Thus, if the plant is being designed with this in mind a
significant development cost can be avoided. This may be true for example, for
the turbine design, the reactor pressure vessel, valves, etc. The economic
evaluation also included an analysis of the fuel cycle cost, using parametric
evaluation of the important parameters. The estimated cost reductions for the
HPLWR compared with a defined reference plant are: 30% reduction for building
and structures, 35% reduction for the reactor plant, 10% reduction for the turbine
plant, and 20 to 25% reduction in overnight capital cost. An initial economic
target for the HPLWR is set at 1000 €/kWe and 3-4 cent/kWh levelized
generation cost.
The HPLWR project was managed under WP VI that included the following
activities: conducting five general project meetings, issuing the Minutes of the
meetings, preparing the project annual report, TIP and final report, reviewing and
editing all deliverable reports, preparing cost and management reports,
communicating with all the partners and with the Commission and maintaining
the project schedule.
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1. Work Package I – Evaluation of current status and
definition of basic requirements
D. Bittermann(Framatome ANP), D. Squarer(FZK), P. Dumaz(CEA), R. KyrkiRajamaki(VTT), C. Maraczy(KFKI), A. Souyri(EdF), N. Aksan(PSI), Y. Oka(U of Tokyo)
1.1
-
1.2
-
-
-
1.3
Objectives
to produce a state of the art report using existing references from partner No. 8 (U. of
Tokyo) and other sources on designs based on supercritical water conditions
to evaluate this state of the art in terms of plant efficiency, core design and technical
difficulties
to identify preliminary design requirements and to define important goals and
conditions for the plant architecture
Description of Work
Study existing literature on design of LWR´s based on supercritical water conditions.
Evaluate the results in general and define the most promising concept and plant
architecture. As main source the work already elaborated by Partner No 8 (U. of
Tokyo) will be considered
Identify and describe the state of the art of supercritical fossil plants including
references and the potential to use turbine technology and other balance of plant
equipment in the HPLWR
Identify essential plant data and conditions like reactor power including the potential
range, fuel cycle, basic architecture, essentials of safety concept
Deliverables and milestones
HPLWR-D1 [1.1], HPLWR-D2 [1.2], HPLWR-D3 [13]
1.4
Methodology
The review work performed under this work package was based on the papers elaborated
by Prof. Y. Oka and co-workers while the definition of requirements and the plant
architecture was based on requirements which are actually valid in Europe for future
reactor types. These are for instance the European Utility Requirements (EUR) and
specific passive design characteristics of the boiling water reactor SWR 1000. In order to
define the core design we have examined in some details the feasibility of the fuel
assembly design of a “reference design” of a supercritical water-cooled reactor that was
studied by K. Dobashi et al.[1.4, 1.7]. This examination have led to several essential
modifications in the “reference design” referred to in WP II. In addition, the experience
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of designing current generations of PWR and BWR was utilized to define design
constraints on the HPLWR [1.5].
1.5
Review of other concepts
The concept of an LWR operating at supercritical conditions has been studied in the past
by different vendors. A review of the various supercritical concepts that were proposed
during the last four decades was elaborated and published by Y. Oka [1.6] and is
described in more detail in D1 [1.1]. Examination of this review indicates that, except for
the University of Tokyo´s SCLWR-H and the CANDU-X reactor designs, none of the
other proposed concepts are likely to be economically competitive with modern LWRs
and therefore are less likely to be developed into a commercial product. Furthermore, the
University of Tokyo´s SCLWR reactor has an additional advantage, in that it can be
designed as an epithermal and a fast reactor (albeit with a low breeding ratio) that could
be fueled with MOX fuel at an enrichment up to 12%, or as a breeder reactor (with
negligible breeding). This is of interest for plutonium management as well as for the
future development of breeder reactors that is required to sustain the nuclear option. We
note here that the original HPLWR proposal to the Commission included an examination
of a breeder design, however this option was deleted from the HPLWR project in order to
make it compatible with the allocated Commission’s funding.
Obvious simplifications and compatibility with LWRs in addition to the higher
temperature raise the possibility of potential cost benefits of this design compared to
existing nuclear power plants. A summary of expected benefits of the HPLWR [1.7] is as
follows:
 Simple plant and reactor system without re-circulation, steam water separation system
of BWR, without steam generator, pressurizer and primary piping of PWR; compact
reactor and plant system
 Applicability of LWR safety principles and basic safety guidelines
 Utilization of advantages of supercritical water coolant by once-through cycle, such
as higher enthalpy rise in the core, low coolant flow rate and higher thermal
efficiency than indirect cycle
 Potential for utilization of LWR technology basis such as RPV, containment, fuel
assembly, control rods, engineered safety features
 Utilization of balance of plant technologies of supercritical fossil power plants such
as turbines, feed-water pumps, feed-water heaters and water cleanup system
A review of the proposal of Oka´s concept [1.4, 1.8] has been performed in order to have
a starting point for further proposals by the partners involved. As criteria for this review
the currently applied requirements and design measures for advanced reactors have been
applied. The major comments on the Oka’s concept [1.4] to be considered are as follows:
 The safety concepts rely only on active safety systems; no passive components are
implemented
 Low reliability is to be expected for steam turbine driven auxiliary feedwater pumps;
complicated and expensive steam line arrangements
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Potential for higher tritium release in case of use of stainless steel fuel cladding
instead of zircalloy
Fuel assemblies can not be characterized as “full utilization of LWR technologies”, as
claimed by Y. Oka [1.4, 1.6] since materials, temperatures and the design features
differ substantially from LWR experience
Moderation concept with “down-flow moderator rods” with insulation lead to very
complicated fuel assembly structure
Currently no manufacturer exists who is licensed to produce fuel rods with
enrichments higher than 5%
Core bypass flow resulting from the core arrangement may lead to significant
reduction of the core outlet temperature and consequently to a reduction in the
expected plant efficiency
According to a Siemens internal report a number of nuclear power plants (17 PWR and
13 BWR) that used stainless steel fuel cladding in the past showed a favorable experience
in PWR but unfavorable experience in BWR due to irradiation assisted stress corrosion
cracking. Stainless steel cladding that contains Co-60 may contribute to a high release of
Co-60 in the primary system and to a higher tritium release (~50% of the generated
tritium) compared with a release from zircaloy cladding (<1% of the generated tritium).
Current licensing target in Germany for tritium release from stainless steel cladding is
expected to be exceeded by a factor of 10-20 and may cause licensing problems. For
LOCA, stainless steel cladding results in about one order of magnitude lower potential
damage to the core than zircaloy cladding; during hypothetical severe accident the
hydrogen release from stainless steel cladding is lower compared with zircaloy cladding;
there are no obvious issues related to fuel storage and transportation; no problems are
expected with the PUREX reprocessing; consequences of higher cobalt doses will have to
be analyzed.
1.6
Design requirements
Since the HPLWR has to be considered as a long term development project, the
requirements applied for the design are expected to be a combination of existing ones
(like EUR) and such discussed for future designs (like Generation IV reactors). This
means among others that in such a design, passive means and means for mitigation of
severe accidents have to be incorporated.
1.7
Proposed design
Due to the very limited funding available for such a task in the work program only very
rough ideas and sketches can be generated. The proposal is to use as a general basis the
concept referenced by Prof. Oka [1.4] and additionally introduce features which consider
the above listed comments, as well as other relevant design requirements that European
vendors such as Siemens and Framatome have been using for many years in the design of
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current generation and advanced LWRs. Some ideas for a plant configuration proposal
are described as follows:
It is desirable to determine reasonable plant sizes depending on the criteria for maximum
RPV diameter and largest commercially available full speed turbine. Depending on the
reactor inlet and outlet temperatures, the maximum practical dimension of the RPV can
contain a core that can yield ~ 2000 MWe, whereas the capacity constraint of the full
speed HP turbine of existing design and without any further modification (800 kg/s at 27
MPa or 850 kg/s at 25 MPa) yields a plant size of ~700 - 900 MWe. Such a plant size
would utilize existing technology and off-the-shelve equipment and therefore is expected
to yield the most economical plant. On the other hand, a plant that is designed for multipurpose applications (e.g. electricity and heat, etc) or for a specific customer may have
different scale considerations. Yet with specific redesign of the turbine, other size
requirements e.g. 1000 MWe can be matched.
1.7.1 Primary System
Framatome ANP has outlined drawings of the core configuration and arrangement. These
drawings contain core and RPV dimensions, flow path, fuel assembly arrangements
(example in Figure 1.1) within the core barrel, mechanical details of the connections and
seals between the fuel assembly, control rods, water rods, tie plates and upper and lower
core plates. In order to avoid temperature deviations in the flange area of the RPV, the
coolant exit pipes were positioned within the center of the coolant inlet. Thus the
complete flange can be held at core inlet temperature (see Figure 1.2). Since the RPV
closure is also at inlet temperature, the leak tightness of the seals is assured. The coolant
flow which is needed for closure cooling is in all cases routed downward through the hot
box and the fuel assemblies to the inlet part of the assemblies by utilization of control rod
guide tubes and moderator tubes. The reflector needs also an intensive cooling in order to
avoid azimuthal temperature differences. This coolant flow enters the transition pieces
which provide the support structure of fuel assemblies. A design of said transition pieces
as individual jet pumps facilitates this flow scheme.
Concerning the fuel assembly it is suggested that the fuel assembly presented in the
“reference design” have tolerance problems as well as other shortcomings as follows: the
grid itself poses new challenges since there is little experience with hexagonal grid;
tolerances for bulging and bending of the fuel assembly must be taken into account; the
fuel pin diameter is too small and may lead to non-economical fuel cycle; the fuel
assemblies differ substantially from current LWR fuel assemblies; there is no obvious
preference for the moderator flow concept; moderation concepts lead to complicated fuel
assembly structures; it is desirable to increase the spacing between fuel assemblies and
move the water rods from the periphery to the inside by one row. This is not expected to
have a large impact on moderation.
The drawings include a detailed sketch of a proposed fuel assembly. Compared with the
“reference design” that has in each hexagonal fuel assembly 258 fuel rods, 30 water rods
and 9 control rods, the proposed hexagonal fuel assembly in a 1000 MWe HPLWR
would have 259 fuel rods, 24 water rods and 6 control rods, a fuel pin diameter of 8 mm,
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a fuel pin pitch of 9.5 mm, face to face channel box dimension of 212 mm, 121 fuel
assemblies, RPV inside diameter of 3380 mm and an annular gap between the RPV and
the core barrel of 200 mm.
Three types of hexagonal fuel assemblies were proposed:
1. Fuel assembly with enlarged water gap between adjacent assemblies and internal
moderator tubes.
2. Fuel assembly with enlarged water gap between adjacent assemblies and solid
moderator rods, control rod guide tubes filled with down-flow water (from closure
cooling).
3. Fuel assembly with minimized water gap (space for handling and tolerances only)
between adjacent assemblies and down-flow water filled control rod guide tubes.
In addition to the hexagonal type fuel assembly (Figure 1.1) also proposals for quadratic
type assemblies have been made.
A rough analysis has been performed on the issue of flow leakages within the RPV. The
flow leakage is caused by differential pressure and temperature, manufacturing
tolerances, and design requirement to remove the heat in the heavy reflector. The analysis
indicates that the by-pass flow may exceed 10% of the total flow for a pressure drop
along the fuel assembly of 1.5 bar. Depending on the differential pressure available to
drive the bypass flow leakage, the reduction in the fluid temperature due to the bypass
flow leakage may reach ~50 ºC and the corresponding reduction in plant efficiency for
the “reference design” would be ~1 % to 1.5 %. Thus, it is necessary to consider the issue
of by-pass flow before finalizing the core design and flow paths.
1.7.2 Containment and Safety concept
Concerning the containment design, the proposal is to consider a modern BWR
containment and in order to introduce passive features into the design, the containment
should be provided with a core flooding pool and emergency condensers (Figure 1.3).
The safety systems configuration should be modified in two essential areas: (1) the
auxiliary feedwater pumps should be omitted, (2) a passive core flooding pool with
emergency condensers should be provided. The reason for this modification is that the
mass of water within the RPV is about 1/10 of that of a BWR or PWR. Therefore, in case
of transients like “loss of offsite power”, the safety philosophy of such reactors to
maintain the primary pressure and consequently the heat capacity of the existing water is
not appropriate in case of an HPLWR. Instead of using a high pressure feedwater
injection with all the complications mentioned above, it is proposed to initiate the
automatic depressurization system (ADS) and use the existing low pressure injection
system and the passive flooding system.
With this proposed containment and the safety concept it is considered that an extensive
use of passive systems is reached by integration of water capacity outside the RPV, that
compensates for the lack of large water mass and natural convection within the RPV of
an HPLWR.
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1.7.3 Balance of plant (BOP)
On basis of the Balance Of Plant (BOP) concept proposed by Prof. Oka essential
differences between steam cycles of fossil fueled plants and the HPLWR are identified.
The most important ones are considered to be that in case of the HPLWR, reheating can
only be performed with steam and if the core outlet steam is used, this corresponds
automatically to a high pressure with negative impact on the design of the reheaters. The
steam conditions after the HP-turbine should be selected in such a way that no moisture
separator is required. Due to the high pressure steam to be used for reheating, the reheater
has to be completely redesigned. Also a redesign of the High Pressure (HP) and
Intermediate Pressure (IP) turbines may be necessary. In addition to the proposed BOP
concept, a feedwater tank is recommended and components and systems for plant start-up
must also be provided.
Steam turbine driven pumps that are incorporated into the “reference design” to cope
with the flow requirement during LOCA are unlikely to meet the European Utility
Requirements (EUR) for LWR nuclear power plants. It is recommended that constant
speed electric motor driven pumps (with a fly-wheel if necessary), with a capacity of
3x50% for a 1000 MWe plant, will be used instead of the steam turbine driven feed-water
pumps; the reliability of steam turbine driven auxiliary feed-water pump is insufficient
due to the following factors: the steam source has limited capacity and variable steam
conditions, it requires sophisticated turbine control system, the steam flow path is
complicated, its reliability is low for startup due to condensate accumulation, it adds
additional high energy steam pipes to the turbine supply, it still requires a small auxiliary
diesel generator.
For the feedwater/steam cycle it is proposed to use a feedwater tank in order to prevent
failure of the feedwater pumps resulting from a failed condenser pump and to change the
heating mode for the intermediate heat exchanger and the last reheater in order to avoid
high system pressure (25 MPa) on the shell side.
On the basis of calculations performed at FZK on the turbine/feedwater circuit, an
evaluation was made on the turbine and reheater assuming an expansion point within the
superheated region after the HP turbine. Concerning a 1000 MWe turbine with full-speed
turbine/generator and use of existing Siemens PG turbine models, either a two-shaft unit
is required or a half speed model must be used.
For the start-up procedure of the plant, the two start-up modes at fixed pressure and at
sliding pressure were evaluated. From economic reasons the result is to propose the
sliding pressure mode, the detailed flow scheme has to be determined later when more
requirements are known. For this operation mode dryout (in the subcritical regime) has to
be considered in the thermal-hydraulics design of the fuel assemblies.
Water chemistry considerations: general requirements are based on the water chemistry
requirements of nuclear power plants and fossil power plants. In PWR, the water
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chemistry of the primary and secondary systems (pH ~9-10 in feed-water) are different
whereas in BWR the addition of hydrogen is partly responsible for the generation of
isotopes of nitrogen and oxygen by radiolysis and this has an impact on plant
maintenance through the ALARA principle. In the Benson boiler oxygenated treatment of
the feedwater has been used in the past in many plants which means that oxygen is added
to the condensate system (pH ~8.5).
It is important to integrate the appropriate water chemistry system into the plant
conceptual design especially with respect to material selection and design of critical
components to be considered, and it is recommended to adopt initially the BWR water
chemistry requirements and systems.
A circuit diagram is proposed that includes components required by the plant startup
system and components that are required for controlling the plant water chemistry (Figure
1.4).
Based on the above mentioned evaluation it is concluded that:
 The selected “reference concept” can serve as a good basis for further HPLWR
development
 No items could be identified that would prevent the feasibility of the concept
 The proposed introduction of passive safety features has to be substantiated by
future development work
 It is assumed that it can be demonstrated that the European Utility Requirements
can be fulfilled (see also Chapter 3 where the results of WP III are summarized).
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Table 1.1- Proposed Characteristics of the HPLWR Power Plant
Characteristics
1
2
Approximate Values
Basis: concept of Prof. Oka incorporating the
following characteristics and modifications:
Gross plant electric output
1000 MWe
Nuclear Steam Supply System (NSSS)
3
4
5
System pressure at the RPV outlet
Fluid temperature at the RPV outlet
Core coolant flow rate
6
Core coolant inlet temperature
25 MPa
500 ºC
To
be
determined
(~1160kg/s)
TBD (~280 ºC)
(TBD)
Reactor Core
7
8
9
10
11
12
13
14
Active core height
Number of fuel assemblies
Average power density
Fuel pin O.D.
Fuel pitch
Fuel assembly shape
Direction of flow in the water rods, however explore
the benefit of solid moderator rods instead of or
together with water rods
Insertion position of control rods
4200 mm
121
TBD by accurate analysis
8 mm (for hexagonal assembly)
9.5 mm
Hexagonal or a square fuel assembly
Downward
Top
Reactor Pressure Vessel (RPV)
15
16
17
18
19
Inside diameter
Design pressure
Design temperature
Wall thickness of cylindrical section
Material
3380 mm
27.5 MPa
350 ºC
300 mm
20MnMoNi55 steel
Containment
20
21
22
Shape
Materials
Pressure suppression system
Cylindrical
Steel reinforced concrete with liner
Drywell, pressure suppression pool,
core flooding pools
Turbine
23
24
Full Speed Supercritical Turbine-Generator
Steam flow rate at turbine inlet
50 1/s
TBD (~960 kg/s)
Decay Heat Removal system (DHR)
25
26
27
28
Accumulators
Steam injectors
Passive core flooding system
Auxiliary feed-water pumps
TBD by accurate LOCA analysis
TBD
Required
Not required
BOP
29
30
31
Feedwater pre-heaters
Heat exchanger/reheater
Feedwater tank
8
After HP turbine
1
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1.8
1.1
1.2
1.3
1.4
1.5
1.6
1.7
1.8
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References
Oka, Y. Koshizuka, S., “D1: Report On Current Status Of Work Concerning
Reactor Designs Under Supercritical Water Conditions”, The University of Tokyo,
HPLWR-D1, April, 2002 (PU)
Bittermann, D.,“Current And Future Turbine Technology Under Supercritical
Water Conditions” Framatome ANP, HPLWR-D2, August 2002 (RE)
Bittermann, D., “General Plant Characteristics”, Framatome ANP,HPLWR-D3,
August 2002 (RE)
Dobashi, K., Oka, Y., Koshizuka, S., “Conceptual Design Of A high Temperature
Power Reactor Cooled And Moderated By Supercritical Light Water”, ICONE 6,
May 1998, ASME, NY, NY
Dumaz, P., “Contribution to the Analysis of Core Design Constraints in
Supercritical- PressureLight Water Reactors”, CEA Technical Note NT-SERILFEA-01, (HPLWR- D15) May, 2001
Oka, Y., “Review of High Temperature Water and Steam Cooled Reactor
Concepts”, HPLWR-SR01,The University of Tokyo, October, 2000 (PU)
Squarer, D., “Reasons for selecting the University of Tokyo Reactor design as a
study reference” CONTRACT N° FIKI-CT-2000-0003, Forschungszentrum
Karlsruhe, December 2000 (RE)
Squarer, D., “Minutes of the second HPLWR project meeting of March 5-6, 2001
at CEA-Cadarache, France” CONTRACT N° FIKI-CT-2000-0003, HPLWRM04, Forschungszentrum, Karlsruhe, April 2001
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Figure 1. 1 -Example of Hexagonal Fuel Assemblies
Page 16 of 62
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Figure 1. 2- Reactor Pressure Vessel (RPV) and Arrangement For Inlet and
Outlet Nozzles
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Figure 1. 3- HPLWR Containment for a 1000 MWe Plant
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Figure 1. 4- Schematic of the HPLWR Circuit Diagram
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Work Package II- Core design and theoretical analyses
G. Rimpault(CEA), P. Dumaz(CEA), C. Maraczy(KFKI), A. Tanskanen(VTT), F.
Wasastjerna(VTT), R. Kyrki-Rajamaki(VTT), Y. Oka(U. of Tokyo), S. Koshizuka (U. of
Tokyo), C. Broeders(FZK), A. Bergeron(CEA), E. Kiefhaber(FZK), D. Struwe(FZK), P.
Rau(Framatome ANP), X. Cheng(FZK),T. Schulenberg(FZK), V. Sanchez(FZK)
2.1
Objectives
The objectives of Work Package II are:
 To evaluate existing core design proposals (i.e., University of Tokyo) including
the results of work obtained from studies on high conversion LWR concepts (i.e.,
French RSM, German PWHCR) related to applicability for the HPLWR
 To propose a preliminary concept for a fuel assembly and control rods
 To identify code requirements for neutronics and thermal-hydraulics for the
HPLWR with thermal spectrum
2.2
Description of Work
The tasks in this work package are:
 Evaluate existing fuel rod and fuel assembly designs considering the requirements
for a HPLWR core
 Make a preliminary estimate of the reactivity coefficients
 Propose a preliminary concept for fuel rod, fuel assembly and control rod
 Evaluate reactor physics and thermal-hydraulics methodologies and cross section
data base applicable to the HPLWR
 Review relevant lattice experiments and assess optimized new experiments that
could be performed (e.g. at PSI’s PROTEUS facility)
2.3
Deliverables and Milestones
HPLWR-D4 [2.10]
HPLWR-D5 [2.1]
The milestones and expected results are:
 Preliminary determination of HPLWR core design and applicable coupled
neutronics/thermal-hydraulics computer codes
 Concept of fuel assembly and control rod
 Deliverables D4, D5
2.4
Methodology
The critical pressure of water is 22.1 MPa and pressure beyond this value is called
supercritical-pressure. Under supercritical-pressure, water does not exhibit a
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discontinuous change of phase and the heat is effectively removed at the pseudo-critical
temperature, which corresponds to the boiling point under sub-critical pressure. The
pseudo-critical temperature of water at the operating pressure of a Super Critical Reactor
(SCR) (25 MPa) is about 385ºC. As there is no discontinuous change of phase under
supercritical-pressure, the SCR does not require steam-water separators. The coolant is
directly fed to the turbine, and therefore the steam generators are not required. The
advantages of a SCR are its compactness, simplicity and the fact that the technologies
developed for Fossil Power Plants (FPP) and Light Water Reactors (LWR) may be used.
Furthermore, the supercritical-pressure light water reactors present some advantages over
existing reactors (PWR and BWR) which prompted the European community (within the
5th FP) to study such a possibility. Any type of spectrum, fast or thermal, could be
envisaged but the 5th FP has limited the task to thermal systems. In order to achieve a
thermal spectrum through the core, some additional moderating materials should be
introduced into the core. Among the various suggested solutions, the one with descending
water in specific insulated water rods has been studied due to its potentially attractive
features.
That solution has been chosen in the so-called Oka’s concept [2.2], which has been
designated as a “reference design” for the purpose of evaluating existing codes and data
suitable for designing HPLWR concepts and as a starting point for further design studies
[2.3].
2.5
Results of benchmark problem
2.5.1
2-D subassembly calculations
Comparison with Monte Carlo code calculations was done [2.1] as follows in order to
verify the calculations: Keff, reaction rate and power for a 2-D slice of the subassembly
at a given elevation. The results obtained with MCNP and ENDFB6 or JEF2.2 cross
section libraries, and with the deterministic codes MULTICELL with ENDFB6 used by
KFKI, ERANOS with JEF2.2 used by CEA, and KARPOS/KARBUS with JEF2.2 used
by FZK were compared. During the performance of these analyses CEA has found that
Hydrogen cross section data are tabulated only up to 350 ºC, therefore there is a need for
Hydrogen scattering test data above 350 ºC. Furthermore, bound Hydrogen data are
necessary for the entire temperature range. If ZrH1.8 is used as a solid moderator, the
corresponding bound hydrogen data should be used. Additional observations are: MCNP
and deterministic codes use Hydrogen temperature-dependent cross sections differently;
MCNP and the deterministic codes behave differently for cases with hydrogen at
temperatures higher than 350 ºC; MCNP uses the highest tabulated cross sections for
cases with higher temperatures; both ERANOS and MULTICELL deterministic codes
use an extrapolation methodology which gives reasonable results but may be unreliable;
the correct method is to use an interpolation technique on α and β before getting the S(α,
β) value which is used to get the thermal matrices; experimental validation to check that
this is correct is not available above 550 ºK; significant differences exist between the
results of calculations using different cross section data sets; the benchmark analysis of
the HPLWR sub-assembly demonstrates that the nuclear data need significant
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improvements before making accurate calculations of the final design; nevertheless, the
current prediction capabilities of the nuclear data and code systems is adequate for
preliminary design.
The following comments regarding nuclear data experiments can be made:
Existing experiments for assessing the nuclear data and computer codes are available for:
PWR or BWR UO2 fuel; different moderation ratios; different parameters (initial
reactivity, power distribution, material balance in burnup calculations, control rod
reactivity worth, feedback reactivity worth, i.e. Doppler, temperature, void, and kinetic
parameters). Current analyses (of experiments in EOLE and PROTEUS facilities) show
that JEF2.2 over-estimate the experimental results whereas ENDFB6 underestimate them.
The conclusions are: the HPLWR benchmark study [2.1] demonstrates that the codes
give reasonable results but the nuclear data need improvements before making the final
calculations; experiments should be performed at high temperatures (350 ºC to 600 ºC) to
determine the impact of bound effects of Hydrogen within water or other hydride
moderator materials; Hydrogen tabulation above 350 ºC should be produced, whatever
the experimental situation for bound effects is; study of the predictability of the nuclear
data and code systems should be done on existing experiments; experiments should be
performed on the final sub-assembly design.
KFKI performed calculations with the KARATE code system, where the effect of Gd
burnable absorbers on reactivity swing was analyzed by performing sub-assembly burnup
calculations of the “reference design” with the MULTICELL code. The analysis was
done on the same six axial assembly slices that were defined earlier by the benchmark
problem of WP II [2.4]. The burnup calculations were presented as a function of Full
Effective Power Days (FEPD) and a 440 day equilibrium cycle. The results of these
calculations indicate that for the “reference design” the burnout of Gd overcompensates
the reactivity loss of fuel burnup and therefore a smaller number of Gd rods should be
used. As a rough estimate 4% of reactivity loss should be compensated between Middle
Of Cycle (MOC) and End Of Cycle (EOC) when the Gd content is unchanged. The
analysis also included calculations with and without Gd, and with the ratio of water rods
flow rate to the total flow rate varying in the range of 0.05-0.60, while the feedback
between neutronics and thermal-hydraulics was calculated with the SPROD code of the
University of Tokyo. This analysis yielded the axial distributions of temperature and
density in the coolant, water rod and wrapper, as well as the fuel temperature and linear
heat rate. Axial peaking factors in the sub-assembly and the effect of control rods on Kinf
(with MCNP) were also calculated. This analysis shows that the axial power profile does
not have a cosine shape but rather a shape with two peaks, the location of which changes
with the flow rate in the water rods and with the introduction of Gd rods (54 Gd rods
were used in the analysis).
The FZK analysis was done for the fuel assembly of the “reference design” with MCNP
models for a 30 º symmetry (similar to VTT and KFKI), full 360 º 2-D and 3-D FZK
models, simplified 2-D and 3-D cylindrical super-cell models for KARBUS, as well as
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coupled neutronics/thermalhydraulics codes: (a) KARBUS multi-group (R-Z) fuel
assembly model and TWODANT transport code and (b) RELAP5 system code with one
channel for reactor core. The results of FZK’s calculations of the benchmark problem
(water density in the range of 0.1-0.7 g/cm3 ) are in general agreement with VTT, while
KFKI and CEA results yield somewhat higher values of Kinf. Significantly higher Kinf
was calculated when the wrapper around the moderator rods was assumed to contain cold
instead of hot water.
2.5.2
Core calculations
FZK did some calculations with 12 and 69 energy groups. Conclusions from this analysis
are as follows: water temperature and density in the wrapper zone have large impact on
the reactivity results for 2-D super-cell slice calculations; 3-D super-cell calculations with
three axial mean fuel enrichments show strong sensitivity to the wrapper treatment (this
was confirmed with MCNP calculations); a good agreement was obtained between the
axial power distributions of the super-cell multi-group and MCNP calculations for the
same densities and temperatures. A striking effect of the water temperature in the
wrapper on the axial power distribution of the sub-assembly was calculated: when the
temperature of the stagnant water in the wrapper was assumed to be as that of the coolant,
the peak of the axial power distribution was calculated at about 1 m from the bottom of
the sub-assembly, but when the temperature of the stagnant water was assumed as that of
the moderator water rod, the peak power shifted to 3.2 m from the bottom of the subassembly. Basically these two axial power shapes are mirror images of each other. Such a
large impact of the wrapper water temperature was unexpected. A verification of the
KARPOS/KARBUS axial power distribution was done by comparing the results against
VTT’s MCNP4C calculations. FZK’s KARPOS/KARBUS, which was validated in the
past for tight lattice LWR, was coupled with FZK’s HPLWR version of RELAP5
containing one channel core representation. Kinf, water density and axial fuel
temperature distribution calculations were performed with these coupled codes for 4, 12
and 69 energy groups and using 1 to 8 iterations between KARPOS and RELAP5. The
results indicate that the 4 energy groups yield significantly lower Kinf, higher water
density and skewed axial fuel temperature distribution as compared with the 12 and 69
energy groups, thus pointing out that at least 12 groups should be used in these
calculations. The conclusions of FZK analysis are [2.1]: convergence of the main
neutronics and thermal-hydraulics parameters is reached within 10 iterations of KARPOS
/ RELAP5; substantially different results are obtained when 12 or 69 energy groups are
used instead of 4 groups; the axial power distribution is very sensitive to the assumed
water temperature within the wrapper; stabilization of the axial power shape seems to be
promoted by temperature changes and Doppler effect on local reactivity; coupled
neutronics/thermal-hydraulics analysis is mandatory; 3-D analysis of both neutronics and
thermal-hydraulics is necessary; the impact of the water inventory between the fuel
assemblies has to be evaluated.
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Because the above mentioned results by FZK may impact some of the conclusions
reached by CEA and KFKI in particular with respect to the: (a) number of energy groups,
(b) effect of water rod and wrapper temperatures and (c) coupled neutronics/thermalhydraulics effects, KFKI did some additional calculations of the “reference assembly”
with the KARATE code system using different group schemes [2.1]. During the HPLWR
core calculations the two energy group structure was used in the KARATE code system.
As the core in the “reference design”[2.2.] is under-moderated and more epithermal than
current LWR, the need for checking the applicability of the two group structure emerged.
Cross-section calculations were performed with the MULTICELL 70 group deterministic
transport code for the “reference fuel assembly” without Gd absorber at zero burnup. The
technological parameter range (coolant density, water rod density, 135Xe concentration, fuel
temperature) corresponded to the nominal conditions. Two, 4 and 6 groups diffusion type
cross-sections were collapsed using the criticality spectrum (B1 equations) for each
parameter combination. The cross-sections were used for the calculation of one
representative assembly. The temperature and equilibrium Xe distributions were calculated
with the KARATE-SPROD code using two energy groups. The SNAP finite difference
code was used to study the effect of the 2, 4 and 6 group scheme, where the cross-sections
were based on the frozen distributions. The results showed that the maximum deviation
between the multiplication factors was about 0.2 %, so the use of the limited number of
groups in this calculation scheme does not lead to serious errors.
It should be pointed out that the results presented by KFKI with respect to the influence
of the number of energy groups differ from the results presented by FZK. FZK found
very significant impact of the number of energy groups while KFKI found little
influence. This is associated to the fact that the group scheme choice depend on the
collapsing procedure of the energy groups, recent codes just like KFKI’s can use broader
group schemes. However, this conclusion stands only for UO2 fuels and an increase in the
number of groups is most probably required for studying MOX fuels.
2.5.3
Conclusions on computer codes for HPLWR
Reference [2.1] summarizes the applicable neutronics/thermal-hydraulics codes, cross
section date bases and neutron physics test requirements. Analyses were carried out by
CEA, KFKI, VTT and FZK. The calculations included: 2-D sub-assembly cell
calculations, 3-D hexagonal core calculation for a limited number of groups and
including burnup calculations, thermal hydraulics calculations including iterations with 3D hexagonal core calculation, reactivity changes over time with burnup and reactivity
coefficients calculations. For the 2-D subassembly analysis the agreement between
CEA’s MCNP, VTT’s MCNP, KFKI’s MULTICELL calculations, and FZK’s MCNP
and KARPOS/KARBUS of k-infinity was within 2% which is significant (the nuclear
data are the main source of these difference). However, such differences exist also when
calculating PWR and the need for significant improvements of the nuclear data is shared
by the reactor physics community, for various reactor applications. More specifically to
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the HPLWR, data on bound hydrogen in water do not exist above 350 ºC in current
libraries. Therefore it should be generated by theoretical analysis and validated by
specific measurements at the ILL facility in Grenoble (France).
KFKI has performed extensive analysis of the HPLWR core while varying the ratio of
water rods flow to the total flow between 0.05 and 0.60 and with
neutronics/thermalhydraulics coupling using the University of Tokyo’s SPROD code.
This analysis showed that the axial power profile does not follow a cosine profile but
rather has a shape with two peaks, the location of which changes with the flow rate.
Based on this analysis, it is recommended that once the fuel assembly design has been
fixed, a core benchmark analysis be carried out together with a parametric and sensitivity
study. With respect to experimental validation of the neutronics analyses, there are some
existing relevant data at the beginning of life (BOL) from previous experimental
programs (e.g. EOLE, VENUS, PROTEUS, etc.) for both UO2 and for MOX as well as
irradiation data.
After the fuel assembly design has been finalized, it is recommended to perform
validation experiments, in particular power map distribution and reactivity worth. The
experiments could determine the pin power, reaction rate and relative reactivity effects on
neutron absorbers in a mockup fuel assembly of the HPLWR and could be performed in
zero power facilities such as PROTEUS or EOLE.
It was concluded that the available analytical tools are adequate for pre-design studies,
however they must be improved for a more advanced and detailed design. Improvements
can be achieved by comparing codes and data, comparing data sets, experiments to be
analyzed and additional experiments to be performed. Also, it is necessary to use coupled
neutronics/thermal-hydraulics codes because of the strong coupling between neutronics
and thermal-hydraulics in a typical HPLWR core.
2.6
Shortcomings And Proposed Modifications To The Fuel Assembly
The original objective of WP II was to propose a preliminary concept of a fuel assembly
(FA), however at present due to the lack of definite proposals on the mechanical
arrangements and the different water flow ranges only potential solutions and direction of
future investigations have been proposed. After evaluating the “reference design” of the
University of Tokyo [2.2] (hexagonal FA) it was concluded that the core is undermoderated and uses excessive neutron-capturing structural material (Ni-based alloy) as
well as non-uniform sub-channel flow, thus imposing a substantial penalty on this
concept. The under-moderation leads to a large reactivity swing that should be
compensated for and the fuel enrichment can reach 7%, which is well above currently
operating commercial fuel production facilities. Also, the different enrichment zones of
the fuel pins within the sub-assembly lead to some complications and the original design
burn-up of 45 GWd/t is too low (proposal is to increase it to 60 GWd/t in order to meet
the EUR and to make the fuel cycle competitive with advanced LWR). The evaluation
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indicated [2.1] that a strong coupling between neutronics and thermal-hydraulics exists
and that additional effort is needed to evaluate MOX fuel cores, plutonium use in the
core, and fast cores. Based on the results of parametric studies for the evaluation of
various options, the following guidelines for improved core design were made [2.5]:
change the structural material of the cladding and the wrapper from Ni-base alloy to
stainless steel thus reducing the average fuel enrichment by 0.9% (however, for some
types of stainless steel of lower mechanical strength, completely annealed thicker
cladding may be required); change the moderation by using zirconium hydrides (or YH2)
instead (or in complement) of water rods and increase the water gap between the subassemblies and filling it with cold water. CEA noted [2.5] that if the cladding thickness
has to be 0.7 mm and a corresponding increase in the wrapper design is required, the
overall volume of structural materials will increase by a factor of two and this will
correspond to an increase of enrichment by more than 1% compared to the “reference
design”. Based on the assessment by WP II the following guidelines for an improved core
were made[2.1], [2.5]: keep the downward water flow-since this helps flatten the axial
power shape and it helps compensate the large reactivity swing caused by undermoderation and the large volume of absorbing material; increase the moderation ratio in
order to reduce enrichment and the reactivity swing; finalize the mechanical design and
flow paths of the fuel assembly; reduce neutron absorption by structural materials; make
the sub-channel flow more uniform.
A series of different actions were performed on [2.4], [2.5]:

Design criteria

Cladding materials

Fuel assembly proposals

Solid moderator
2.6.1
Design criteria
With respect to some of the design constraint criteria, it was suggested [2.6], [2.7] to use
a nominal cladding temperature of 620 ºC, a maximum cladding temperature of 1200 ºC
for class 3 and 4 transients, a maximum fuel temperature of 1930 ºC and a low value of
linear heat rate of 270 W/cm. It should be noted that this value of the linear heat rate is in
dispute since the University of Tokyo has used a value of 390 W/cm. A more accurate
design constraints values could only be determined by performing a thorough transient
and safety analysis of the HPLWR.
2.6.2
Cladding materials
VTT did some neutronics calculations [2.5], [2.7] for the HPLWR cladding material with
the objective to find a material that will improve the neutron economy and has minimal
radiation damage.
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Calculations were carried out [2.8] with MCNP using cross sections of nickel and iron in
steel from IRDF-90 version 2 and assuming the conditions of WP II benchmark case No.
4 with the neutron source based on the average power density. The calculated
displacement rates were multiplied by the total peaking factor (2.26) of the “reference
design” to obtain the maximum cladding damage rate. This analysis resulted in a
maximum nickel and iron displacement damage rates of 4.2*10-7 and 3.8*10-7
displacement per atom per second (dpa/s) respectively. The same data and conditions
were used for the calculation of the maximum helium production rates in Inconel 718 and
alloy 1.4970. The results of these calculations [2.8] indicated that steel cladding absorbs
less neutrons than Inconel cladding (Keff = 1.09750 and 1.14811 for Iconel 718 and alloy
1.4970 respectively). Dominant helium production processes in these materials are
10
B(n,α)7Li and 58Ni(n,γ)59Ni(n,α)56Fe.
The calculated helium production rates for Inconel 718 and alloy 1.4970 are <2.50*10 -7
and <7.04*10-8 atomic parts per million per second (appm/s) respectively, which amounts
to approximately 87 appm He after three full power years. This analysis indicates that He
production in these cladding materials may be minimized by minimizing the boron
content in the cladding [2.5], [2.7]. [2.8].
Framatome ANP designed the RPV internals and estimated the flow distribution as being
approximately 1/3 through the downcomer, 1/3 through the reflector and 1/3 through the
water rods [2.5]. The RPV internals include a hot box above the core which is a
mandatory feature of the design. Furthermore it is important to have a uniform
temperature distribution within the reflector (within ±2 ºC) in order to avoid structural
deformation of the reflector [2.5].
2.6.3
Fuel assembly proposals
Framatome ANP suggested two fuel assembly (FA) designs [2.4], [2.5], [2.7], [2.9]. The
square geometry has been designated as a “wet” FA and the hexagonal one as a “dry”, to
reflect the larger volume of moderator water in a square FA. A square fuel assembly is
also advocated by FZK [2.7]. The following geometrical data pertain to each FA: Square
FA- 165 fuel rods, total area 488 cm2, 25 square moderator/guide tubes, moderator
annulus 58 cm2 , moderator tube area 183 cm2, total moderator area 241 cm2, flow area
37 cm2, total water area 278 cm2. For the hexagonal FA – 435 fuel rods, total area 397
cm2, 34 round small moderator/guide tubes, moderator annulus 7.2 cm2, moderator tube
area 17.5 cm2, flow area 128 cm2, total flow area 153 cm2. For the hexagonal FA it was
assumed that the gap between two FAs can not be less that 2 mm due to tolerances,
refueling and lateral support. Comparison between the two fuel assemblies indicates a
preference towards a square fuel assembly in order to obtain a high and more uniform
reactor exit temperature. Double wall moderator tubes as well as the FA outer structure
are complicated and expensive and their feasibility may be in doubt.
2.6.4
Solid moderator
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ZrH1.8 has less moderation by approximately 20% as compared to water. Thus, if solid
moderator will be used in the design of the HPLWR it will necessitate a 20% larger
volume compared with water. Extending the burnup to 60 GWd/MTU will require an
increase in the enrichment by ~0.5 % and that the use of Ni alloys as a cladding material
will cause an increase in the enrichment by ~1%.
FZK did the analysis of a square fuel assembly. Obviously, the conclusions of this analysis
depend on the assumed geometry and materials of the fuel assembly. Results were
presented [2.5] for Kinf vs. enrichment for a square and hexagonal assemblies, Kinf vs.
water density for both fuel assemblies and power distribution and peaking factor for a
single enrichment. The following conclusions were drawn as a result of FZK’s neutronics
analysis [2.5]: deterministic multi-group neutronics models are required for the calculation
of burnup and transients; for the “reference assembly” a very strong coupling between
neutronics and thermal-hydraulics exists; a deterministic super-cell model of the FA, based
on the water rod and its surroundings, could be validated by Monte Carlo simulations; the
first neutron physics analysis for a square FA proposed by FZK indicates the need for
higher enrichment (6.5% vs. 5.1% for the “reference design”) due to thicker Ni-based alloy
assumed in the calculations. Also, an increase in the equivalent core diameter of 5-10%
may be expected; the analysis showed potential problems associated with the water rods
design of the “reference design”, due to the influence of the thermal insulation around the
water rod on core neutronics and core thermal-hydraulics, as well as due to the practical
aspects of constructing such a FA. Alternatively, solid moderator rods could be used for
example in the upper part of the core in selected pins in order to compensate for the lower
coolant density of the hot coolant. Such a FA would be similar to the FA of an epithermal
High Conversion LWR with MOX fuel that was studied by FZK in the 1980’s.
2.7
Conclusions
The conclusion after performing the analysis reported in reference [2.10] is that it is
believed that with appropriate design changes, the HPLWR core can be improved
substantially by addressing the above mentioned issues.
2.8
2.1
2.2
2.3
References
Rimpault, G. et.al. Applicable neutronics/thermal-hydraulics codes, cross section
data base and neutronics test requirements, HPLWR-D5, September 2002
Dobashi,K, Oka, Y., Koshizuka, S. .: Conceptual Design of a High Temperature
Power Reactor Cooled and Moderated by Supercritical Light Water, Ann. Nucl.
Energy, vol.25, pp.487-505 (1998) (also ICONE-6, May 10-15, 1998)
Squarer, D., “Reasons For Selecting The University Of Tokyo Reactor Design As
A Study Reference”, Forschungszentrum Karlsruhe, Germany, HPLWR-MEM01,
December, 2000 (RE)
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Squarer, D.,Bittermann, D.,Oka, Y.,Dumaz, P.,Rimpault, G.,Kyrki-Rajamaki, R.,
Ehrlich, K.,Aksan, N.,Maraczy, C.,Souyri, A., “HPLWR Annual Technical
Report”, HPLWR-D12, September 2001 (RE)
2.5
Squarer, D. “Minutes of The Fifth HPLWR Project Meeting of July 29-31, 2002
at FZK – Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M 11, September 2002 (RE)
2.6
Dumaz, P., “Contribution to the Analysis of Core Design Constraints in
Supercritical-Pressure Light Water Reactors”, CEA Technical Note NT-SERILFEA-01, May, 2001, HPLWR-D15 (RE)
2.7
Squarer, D. “Minutes of The Fourth HPLWR Project Meeting of March 4-6, 2002
at EdF – Chatou, Paris, France”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M 10, April 2002 (RE)
2.8
Ehrlich,K., Konys,J, Heikinheimo, L., Leistikow, L.,Steiner, H., Arnoux, P.,
Schirra, M., „ In-core and Out-of-core Materials Selection for the HPLWR”,
Forschungszentrum Karlsruhe, Germany, HPLWR-D8, August 2002 (RE)
2.9
Bittermann, D. “General Plant Characteristics”, Framatome ANP, Germany,
HPLWR-D 03, August, 2002 (RE)
2.10 Rimpault, G. et.al.,”Results of Evaluation of Existing Core Designs and Potential
for Application for HPLWR”, HPLWR-D4, September 2002
HPLWR – D 13
3.
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Work Package III – Reactor Safety and Deteriorated
Heat Flux
N. Aksan(PSI), D. Bittermann(Framatome ANP), D. Squarer(FZK), T.
Schulenberg(FZK), X. Cheng(FZK), D. Struwe(FZK), V. Sanchez(FZK), P. Dumaz(CEA),
R. Kyrki-Rajamaki(VTT), A. Souyri(EdF), Y. Oka(U of Tokyo), S. Koshizuka(U of Tokyo)
3.1
Objectives
Preliminary assessment of prospective safety features of the thermal HPLWR with the
ultimate goal to ensure that the HPLWR is at least as safe as the latest versions of LWRs
(e.g., EPR, SWR 1000, ABWR, APWR)
Review of the status of the deteriorated heat transfer characteristics and pressure drop in
supercritical water near the pseudo-critical line (similar to Critical Heat Flux (CHF)) at
conditions of relevance to the HPLWR
3.2
Description of Work
- Make a preliminary identification of the required safety features and the safety design
requirements of the HPLWR based on the latest safety philosophy and features of
advanced LWRs
- Identification of passive safety systems that could be incorporated into HPLWR
- Make a preliminary assessment of the appropriateness of the RELAP5 code to perform
thermal-hydraulics LOCA and transient analyses of the HPLWR
- Preliminary evaluation of the need for safety tests in existing large scale facilities
- Review of heat transfer phenomena in supercritical fluids with emphasis on heat
transfer deterioration and pressure drop
- Preliminary identification of potential computer codes (system codes/CFD) and code
requirements that could be used for the design and analyses of CHF tests
3.3
Deliverables and milestones
HPLWR-D6 [3.15]
HPLWR-D7 [3.10]
3.4
Assessment of Required Safety Features
In the “reference design” [3.2] only minimal information is given about the containment
design and the primary system, and the containment concept hardly pays any attention to
passive safety systems. Consequently substantial effort has been invested in order to
define the safety features of the plant in a European environment, as well as to
incorporate passive safety features into the design.
The major basic requirements for safety and licensing are as follows:
 Licensibility in different countries (standardization)
 Consideration of design extension conditions (complex sequences, severe accidents)
and prevention of early containment failure
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





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Application of standard PSA methodology
Integral frequency target for core damage (10-5 per year) and limiting releases target
for severe accident (10-6 per year)
Minimal emergency protective action beyond 800 m from the reactor
No delayed action (temporary evacuation) beyond 3 km
No long-term action (permanent resettlement) beyond 800 m
Limited restriction of consumption of foodstuff and crop in area and time scale
3.4.1 Containment Concept and passive safety features
The containment concept is based on that of recent-generation BWR plants. It is a
cylindrical containment made from steel reinforced concrete equipped with an inner liner
and a pressure suppression system (Figures 1.3 and 3.1). The containment is divided into
a drywell, which includes four flooding pools and a pressure suppression pool, as
required by the pressure suppression system. It is considered that the containment design
pressure is determined by a maximum LOCA, which is expected to result in a design
pressure of about 0,2-0,3 MPa.
3.4.2 Safety Concept
It is the aim for the development of the HPLWR to use both passive and active safety
systems for performing safety-related functions in the event of transients or accidents.
The most frequent events requiring system function for prevention of intolerable fuel rod
temperatures comprise anomalies in plant operation, or so-called transients. As a result of
the specific conditions of supercritical water the water inventory within a HPLWR is
about 1/10th of that of a BWR or a PWR. This means that in case of incidents and
accidents, the heat storage capacity of the existing water is low. Concerning the control of
incidents and accidents this fact has to be considered appropriately. In general this means
that as fast as possible, flow has to be maintained which is able to cool the core. Later on
the core has to be flooded with water from all sources, including water reservoirs external
to the primary circuit.
From the analyses for a hot line break and a loss of feed-water flow accident [3.6], [3.14]
it is expected that the core cooling will be more effective in case of loss of flow
accidents, if the Automatic Depressurization System (ADS) is activated and followed by
a low pressure water injection from the suppression pool, compared to high pressure
injection. Although this has to be substantiated by further analyses, this procedure seems
to be the appropriate mode to control these kinds of accidents.
Therefore in case of incidents with loss of feed-water flow it is proposed (as in the
original Oka design [3.2], [3.9]) to apply the principle of ADS followed by low-pressure
coolant injection (see Table 3.1 and Figure 3.1 below). Whether accumulators can be
used in addition or even instead of the pumps has to be analyzed further. This mode
should result in the lowest temperature loads of the fuel rods and in reliable systems for
accident control. It should be pointed out that the same design philosophy has also been
adopted in the design of Advanced Light Water Reactors (ALWR).
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In case of most of the transients as well as in the event of accidents, the following safety
functions must be assured:






Reactor scram
Containment isolation
RPV pressure relief and depressurization
Heat removal from the RPV
Reactor water makeup and control of core coolant inventory
Heat removal from the containment
The passive and active systems planned for these tasks are described in Table 3.1 below
Table 3.1 – Proposed passive and active safety systems for the HPLWR
Safety functions
Systems provided
Reactivity control
Two independent scram systems
Containment isolation
2 main steam isolation valves per train
Reactor pressure control and reactor
depressurization
6 safety relief valves; 4 emergency
condensers
Core flooding
4 RHR and LPCI systems; flooding lines;
possibly accumulators
RHR from RPV
4 RHR and LPCI systems; 4 emergency
condensers
RHR from containment
4 RHR and LPCI systems; 4 containment
condensers
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1
2
3
4
5
6
7
8
9
10
7
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Safety relief valves
Emergency condenser
Core flooding pool
Pressure suppression pool
Low pressure injection
Drywell flooding line
Containment cooling condenser
Vent pipes
Main steam line
Feedwater line
11
12
13
14
15
16
17
18
19
HP turbine
LP turbine
Condenser
Condensate water pump
Feedwater tank
Main feedwater pump
LP Reheater
HP Reheater
Water separater, reheater
1
9
3
19
2
10
11
12
8
4
4
13
6
5
18
16
15
17
14
HPLWR
Circuit diagram, schematic
Figure 3. 1–
Containment and Primary Circuit Concept for the HPLWR
3.5 General Application of Some Safety Requirements
Since the HPLWR is considered to be a long term development project which is expected
to be realized in the far future (by approximately 2015, similarly to the Generation IV
nuclear reactors that are now being assessed by the U.S. DOE and the Generation IV
International Forum (GIF), it is somewhat difficult to foresee the requirements which will
be appropriate at that time. Therefore it was decided to take into account as a general
guide, the European Utility Requirements (EUR), which are currently considered to be
most advanced and most complete in Europe and have been applied in the design of
advanced LWRs such as the EPR and the SWR 1000 (detailed designs of which are very
advanced). Additionally, the trends of future requirements, as expressed in the
requirements known from the Generation IV initiative, was considered in order to include
further advanced ideas.
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3.5.1 European Utility Requirements for Safety versus HPLWR
Application of EUR safety requirements to HPLWR has been performed as a guideline in
general terms and reference [3.16] has been used as the basis. Additional EUR chapters
(e.g. 2.2, 2.4, 2.8, 2.9) are necessary to carry out specific application and compliance
assessment, which can be performed after having a mature design. Presently, it is
necessary to keep a close overview for the preliminary basic design in relation to the
EUR safety requirements. These requirements have been considered and taken into
account during the evaluation of the merit and feasibility of the HPLWR concept and the
results are provided in tables [3.15] for the primary system and for the containment
system.
A comparison of EUR requirements with the HPLWR general features is also made
[3.15]. The objective of this comparison is to show that we estimate that the HPLWR has
a potential to meet these requirements. However, it should be realized that the task of
comparing the HPLWR to the EUR is quite substantial. Such a comparison can only be
made after the HPLWR design would become more mature and would include adequate
detailed description of the entire power plant. On the other hand, it is advantageous to
examine carefully the EUR requirements during the design stage, in order to assure the
fulfillment of these requirements later on. Thus it is clear that such a process is iterative,
and a design of a new power plant such as the HPLWR can benefit from substantial
savings by considering the requirements in every step from the very beginning.
3.5.2 Generation IV Technology Goals in The Safety and Reliability Area versus
HPLWR
Since many aspects of conceptual HPLWR design are not known in detail until the
completion of the basic design and the related research and development efforts, general
guidelines for the Generation IV technology goals in the safety and reliability area vs.
HPLWR are provided in tables [15]. This comparison indicates that the present
preliminary design of the HPLWR has the potential to meet most of these goals. It is also
to be noted that, in general, the Generation IV requirements are generally compatible
with the top tier EUR document. This is an important observation, since by using the
EUR as a guide for the detailed design of the HPLWR, it will also insure the conformity
of the HPLWR with Generation IV goals.
3.6
Deteriorated Heat Transfer in Supercritical Water
Extensive literature survey was undertaken on heat transfer at super critical conditions
and also on heat transfer deterioration at the pseudo-critical point. The details of this
work are described in [3.10, 3.11]. Considering the valid parameter range of correlations,
it is found that the correlation of Bishop is the most suitable one for the sub-channel
conditions of the HPLWR “reference design” and it was recommended for use in the
HPLWR project. With respect to the onset of heat transfer deterioration, there are no
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experimentally verified correlations applicable to the HPLWR conditions. However, for a
preliminary assessment, the correlation of Yamagata [3.10, 3.11] is recommended.
Thermal physical properties of supercritical water were also reviewed in the same report
[3.11]. It was agreed that the steam table of IAPWS-97, respectively IAPWS-95, was
well accepted by international institutions and, therefore, should be taken as the reference
for the HPLWR project. In addition, recommendation for the friction pressure drop at
supercritical conditions was also made.
3.7
Proposed Design
Since the core design and configuration of the HPLWR has been changing and there is
presently no fixed core configuration (e.g. the fuel assembly shape could be hexagonal or
square, and various versions of these configurations), it was agreed to concentrate on
hexagonal fuel assembly and on the safety systems of the “reference design”. Without
this assumption, it is nearly impossible to progress due to the limited resources. It should
also be noted that the “reference design” has only minimal information about the
containment design and related safety systems. In the mean time, there has been a
proposal by the HPLWR project (see Chapter 1) for the containment and safety concepts
with some details as indicated in Section 3.4 above.
3.8
Preliminary Transient Safety Analyses of the HPLWR
The supercritical HPLWR reactor represents a challenge for best-estimate safety analysis
codes like RELAP5 and CATHARE since such codes were developed for two-phase or
single-phase coolant at pressures far below the critical point. Hence the prediction of the
thermo-physical properties of steam and the available correlations for the wall-tosupercritical water heat transfer has not been validated yet.
The safety analysis codes RELAP5, CATHARE and TRAB (TRAB is also a reactor
dynamics code), which were developed for current generation LWRs, were modified in
order to be able to perform HPLWR safety analyses. Thus, the steady-state conditions
and some selected transients, in which the system pressure remains above the critical
pressure, were predicted by RELAP5, CATHARE and TRAB. The RELAP5 code was
also used to evaluate the impact of loosing the thermal insulation of the water rods and to
study the feasibility of substituting solid moderator rods instead of water rods.
The University of Tokyo has also made available to the HPLWR partners (on bilateral
basis) several computer codes, which were developed for the SCLWR design, e.g. 1-D
steady-state thermal-hydraulic analysis code, 1-D transient code and a LOCA code.
3.8.1 Use of RELAP5/Mod3 Computer Code
Work on the preliminary assessment of the appropriateness of RELAP5 code has
progressed in the direction of two-fold approach [3.12, 3.13]. These two approaches are:

Simple modelling of the HPLWR and an attempt to verify the application of the
RELAP5 code to the selected transients (e.g. LOCA),
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
Selection of relevant phenomena, parameter ranges and transient types and
assessment of the applicability of the RELAP5.
As part of the second approach, a list of transient phenomena for HPLWR conditions has
been identified and selected: These are:
-
Critical flow at super-critical conditions (as also confirmed by the
investigations at FZK)
-
Reflooding for tight lattice geometry at low pressure
Boil-off for tight lattice geometry at low pressure
An attempt to assess the critical flow at super critical conditions was made at PSI using
the Edwards’ pipe blowdown experiment. This experiment simulated a pipe pressurized
with water that was blown down to the atmosphere through a large, fast-opening hole in
one and of the pipe. Although actual tests were performed at sub-critical conditions, the
initial pressure of the simulation was changed to 25 Mpa and also the initial temperature
was changed, for this evaluation. The calculation, which was performed, encountered
water property failures and could not run to completion with the existing RELAP5
version. The failure occurred near the pseudo-critical temperature point. The reasons of
failure are under investigation.
The NEPTUN-III bottom flooding experiments were used to assess the capabilities of
RELAP5/Mod 3.3 to predict reflooding in a tight lattice geometry at low pressure. Since
the experiment and analysis deal with the reflood stage, it includes only the sub-critical
range. The experiments and analysis describe the reflooding phenomena in a tight
hexagonal lattice following hypothetical LOCA scenario. The results indicate that for a
tight lattice the RELAP5 code under-predicts the peak cladding temperature measured
during a series of reflooding experiments performed at the NEPTUN-III heated rod
bundle facility. The reasons for these differences are discussed in [3.17] and potential
improvements in the RELAP5/Mod 3.3 modelling of reflooding in tight-lattice assembly
have been investigated. It should be pointed out that correct modelling of relevant
experimental data (e.g. NEPTUN-III, etc.) by codes such as RELAP5 or other codes is a
necessary step in code verification and often requires extensive effort. Although the final
HPLWR fuel assembly may have a different geometry, i.e. a square fuel assembly, a
similar code verification effort (for other codes, e.g., RELAP5, CARHRE, FLICA or a
sub-channel code) may be required.
At FZK extensive work has been performed within the HPLWR-project to evaluate the
appropriateness of the best-estimate, two-phase flow thermal hydraulic system code
RELAP5 as a reliable tool for safety-oriented investigations as well as supporting studies
during the plant and core optimization process. These activities were mainly focused on:
-
review of physical models, e.g., thermo-physical water properties,
wall/supercritical water heat transfer
-
review of code’s numeric for the prediction of the thermo-physical water
properties e.g. close to the critical point
-
review of code’s simulation capabilities, e.g., for plant steady-state conditions,
plant transients and accidents
identifying problem areas and model needs regarding the HPLWR-peculiarities,
e.g., supercritical pressure
-
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proposing areas where future work must be oriented.
As a continuation of the qualification work, selected plant transients, where the system
pressure remains above the critical pressure, were investigated with RELAP5 aiming to
get a first impression how RELAP5 behaves in the supercritical regime rather than in
assessing the plant response during transients. In addition, some exploratory LOCA
investigations were carried out to examine whether RELAP5 is also able to predict the
HPLWR plant behavior during accidents, where the system pressure decreases below the
critical pressure e.g. during the blow-down phase of loss of coolant accidents.
The following transients were analyzed with RELAP5 at FZK:
1) loss of feed water heating
2) reduction of coolant flow
3) loss of off-site power
The sequence of events, the initial and boundary conditions as well as the assumptions
for the calculations were mainly taken from previous studies performed by the University
of Tokyo [3.2, 3.6-3.8] using codes especially developed for this type of reactor.
In general, it can be stated that RELAP5 has the potential to be used as a reliable safety
analysis tool for the assessment the HPLWR plant. However additional improvements of
numeric and physical models as well as further code qualification work are necessary to
fully cover the analyses of postulated HPLWR transients and accidents.
3.8.2 Use of CATHARE 2 Computer Code
Analyses of HPLWR reactor transients (like LOCAs) require a thermal-hydraulics code
that is able to calculate supercritical fluid flow, two-phase flow in sub-critical conditions
and transitions between these conditions. For such purpose, the capabilities of the
CATHARE2 (version V1.5a) computer code have been investigated at CEA (Cadrache
and Grenoble). CATHARE is usually used for pressures in the sub-critical regime, but
thanks to steam tables that include properties up to 26 MPa, it was possible to develop a
modified version of the standard code, which is able to simulate transients where both
supercritical and sub-critical regimes are encountered.
The “reference concept” of the HPLWR project (SCLWR-H) [3.2] has been modeled
with CATHARE using:
- one-dimensional modules, 1-D (feed-water lines, downcomer, fuel coolant channel,
etc.)
- volume modules, 0-D (lower plenum, upper plenum, etc.)
- boundary condition modules, BC (inlet and outlet conditions)
Two transients were analyzed with the CATHARE code using a very simple model of the
HPLWR. These transients are: feed-water line break and steam line break. In both cases,
the transient behavior appears to be well calculated from supercritical to sub-critical
regimes. In fact, the supercritical regime lasts one or two seconds only. However,
additional effort is necessary to improve the physical models in the code, to verify it and
to validate it against experimental data. The work achieved up to now has shown that
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CATHARE2 has the potential, after further code verification and validation, to be used as
a reliable safety analysis computer tool for HPLWR type reactor.
3.8.3 Use of TRAB Computer Code
As an example the following transients were analyzed with the RELAP5/MOD3 code
[3.15, 3.19]: (1) loss of feed water heating, (2) reduction of coolant flow, and (3) loss of
off-site power. The TRAB code of VTT was used to analyse the following transients
[3.13, 3.18]: (a) inlet temperature transient (280 to 260 ºC in 1 sec), (b) inlet mass flow
transient (1816 to 908 kg/s in 1 sec), (c) outlet pressure increase transient (25 to 27 MPa
in 1 sec), (d) outlet pressure decrease transient (25 to 23 MPa in 1 sec). The results of the
TRAB analyses indicate that the code can, in principle, analyse the HPLWR dynamics
and showed that the HPLWR system behaves as expected at super-critical pressures and
that the transients decay after 10-20 seconds. Additional modelling effort is needed to
improve the code capabilities.
3.8.4 Future work
It is evident from the work summarized above that several fundamental HPLWR-related
issues need to be experimentally and theoretically investigated in more details within a
technological research program. In this context, the research activities related to the
qualification and validation of thermal-hydraulics and safety analysis tools may be
concentrated on the following areas:
1. Fundamental heat transfer mechanisms

Heat transfer for wall/supercritical water under steady-state conditions

Heat flux deterioration for supercritical water

Heat transfer during boil-off and reflood of relevant lattice cores

Critical flow of supercritical water
2. Validation of developed correlations for supercritical water against experimental
data
3. Development of appropriate interpolation schemes to predict the steam/water
properties around the critical point.
4. Steady-state investigations aimed to optimize different moderator rod concepts
5. Coupling of thermal-hydraulics codes with neutronics codes
6. Safety evaluation of proposed plant design with best-estimate codes coupled with 1D or 3-D-kinetics
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7. Supporting investigations to optimize the safety systems design
From the HPLWR investigations to-date it can be concluded that the above mentioned
codes have the potential to be used as reliable safety analysis tools within the framework
of the HPLWR project. However additional improvements of numerical method and
physical models as well as further code qualifications are necessary to fully cover the
analyses of postulated HPLWR transients and accidents.
3.9
Conclusions
At the conclusion of the HPLWR project under the 5th FP, the HPLWR design has not
been completed in sufficient details to allow the performance of accurate safety analysis,
the expected regulations have not been explored and the computer codes that could be
used to perform safety analysis have not been validated and verified for supercritical
water conditions. However, despite these shortcomings, the safety related information
presented in the deliverable reports [3.10, 3.15], supports the contention that the HPLWR
can be designed to operate safely and is expected to reach the safety level of advanced
LWRs. Furthermore, the information presented in the reports [e.g. 3.15, 3.19] clearly
demonstrates that reliable well-known computer codes like RELAP5 and CATHARE that
are currently being used for the safety analyses of LWRs will be able to analyze the
HPLWR after additional development and verification. The preliminary results shown in
these reports also indicate that these codes, in their present preliminary status, can already
be used to help define and optimize the necessary safety systems of the HPLWR.
Moreover RELAP5 can, at this stage of the project, be used for core design optimization
studies.
Since the HPLWR will draw from the experience of existing LWRs and have some
features (e.g. lack of Zr, less fuel, simplified circuit) that increase its safety potential, it is
believed that the HPLWR can be designed and licensed by regulatory authorities to
operate safely.
3.10
References
3.1
V. H. Sánchez, et. al.; Investigations of the Appropriateness of RELAP5 to Analyse the
Safety Features of the HPLWR-Reactor. FZK-Report in preparation
K. Dobashi, Y. Oka, S. Koshzuka, "Conceptual Design of a High Temperature Power
Reactor Cooled And Moderated By Supercritical Light Water", ICONE-6, May 10-15,
1998, ASME, NY.
D. Barber, T. Downar, W. Wang; Final Completion Report for the Coupled
RELAP5/PARCS Code. PU/NE-98-31. 1998. Purdue University.
3.2
3.3
HPLWR – D 13
3.4
3.5
3.6
3.7
3.8
3.9
3.10
3.11
3.12
3.13
3.14
3.15
3.16
3.17
3.18
3.19
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H. Finnemann, R. Böhm, J. Hüsken, R. Müller, J. Mackiewicz; HEXTIME: A hexagonal
space-time kinetics code for the analysis of PWHCR transients
T. Schulenberg, “Minutes of the WP I Meeting of April 27, 2001 At Forschungszentrum
Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany, HPLWR-M 05, May,
2001 (RE)
S. Koshizuka, k. Shimamura, Y. Oka; large-break Loss-Of-Coolant Accident Analysis of
a Direct-Cycle Supercritical-Pressure Light Water Reactor. Ann. Nucl. Energy, Vol.21
No.3, pp 177-187.1994
J. H. Lee, S.Koshizuka, Y. Oka; Development of a LOCA Analysis Code for the
Supercritical-pressure Light Water Cooled Reactors. Ann. Nucl. Enegy Vol. 25. No.16.
pp 1341-1361.1998.
K. Kitoh, S. Koshizuka, Y. Oka; Improvements of Transient Criteria of Supercritical
Water Cooled Reactor Based on Numerical Simulation. ICONE-2341. 1997
Oka, Y., “Review of High Temperature Water and Steam Cooled Reactor Concepts”,
HPLWR-SR01, October, 2000
X. Cheng, T. Schulenberg, A. Souyri, V. Sanchez, N. Aksan, “Heat Transfer and Pressure
Drop in Supercritical Pressure- Literature Review and Application to a HPLWR”,
HPLWR-D 07, September, 2001 (PU)
Cheng, X., Schulenberg, T., “Heat Transfer At Supercritical Pressures-Literature Review
and Application to an HPLWR”, Forschungszentrum Karlsruhe, Technik und Umwelt,
FZKA 6609, May 2001
Aksan, N., Schulenberg, T., Cheng, X., “Minutes of the WP-III Technical Meeting of
January 29-30, 2001, At Forschungszentrum Karlsruhe, Germany”, Paul Scherrer Institut,
Switzerland and Forschungszentrum Karlsruhe, Germany, HPLWR-M 03, February,
2001 (RE)
Squarer, D. “Minutes of The Second HPLWR Project Meeting of March 5-6, 2001 (Rev.
1) at CEA – Cadarache, France”, Forschungszentrum Karlsruhe, Germany, HPLWRM04, April 2001 (RE)
Koshizuka, S.,Oka,Y., “Computational Analysis of Deterioration Phenomena and
Thermal-Hydraulic Design of SCR”, SCR-2000, pp169-179, Nov. 6-8, 2000, Tokyo
N. Aksan, D. Bittermann, D. Squarer, “Potential Safety Features of The HPLWR and
General Application of Some Safety Requirements” EC-Report, HPLWR-D 06,
September 2002 (RE)
EUR, “Volume 2: Generic Requirements, Chapter 1: Safety Requirements (Parts 1 and
2)” Revision B, November 1995 (As supplied by Framatome ANP)
G. Th. Analytis, “Analysis of seven NEPTUN-III (tight lattice) bottom flooding
experiments with RELAP5/MOD3.3/BETA”, PSI Internal Report, TM-42-02-07, July 23,
2002
Leppanen, J, Tanskanen, A., Kyrki-Rajamaki, R., “Feasibility of the TRAB code for
HPLWR reactor dynamics calculations”, VTT Processes, Project Report YR-PR-14/02,
August 2001
V. H. Sánchez, et. al.; Investigations of the Appropriateness of RELAP5 to Analyze the
Safety Features of the HPLWR-Reactor. FZK-6947, 2002.
HPLWR – D 13
4.
HPLWR Contract No. FIKI-CT-2000-00033
Page 41 of 62
Work Package IV-Summary Report on Material Selection
and available Treatments for Reduction of Corrosion in
HPLWR components.
K. Ehrlich (MCS-FZK), J. Konys (FZK), L. Heikinheimo (VTT), S. Leistikow (FZK
Consultant), P. Arnoux (CEA), M. Schirra (FZK)
4.1
Objectives
Perform a state-of-the-art study that will guide in-core and out-of-core materials
selection for the HPLWR.
4.2
Description of work



4.3
Evaluation of existing materials for fuel elements, core structures and piping
and other relevant components based on assumed boundary conditions for a
thermal HPLWR, and preliminary selection of appropriate candidate
materials; Identification of potential future experiments.
Metallurgical characterization of potential materials and optimization towards
reduced corrosion/stress corrosion cracking by thermal-mechanical and
surface/process treatments
Review of the effects of fluid radiolysis and power plant water chemistry on
candidate HPLWR.
Deliverables and Milestones
HPLWR-D8 [4.1], HPLWR-D9/10 [4.2]
In Work Package IV of the HPLWR project a state-of-the-art study was performed to
investigate the operational conditions for in-core and ex-vessel materials in a future High
Performance Light Water Reactor (HPLWR) and to evaluate the potential of existing
structural materials for application in fuel elements, core structures, reactor pressure
vessel (RPV) and out-of-core components [4.1], [4.2].
For the conventional or ex-core components like boilers, superheaters and turbines of
this novel plant the operational design data, summarized in Table 4-1, are moderate with
regard to the expected temperature ( 600°C) and pressure levels (250-275 bar) of the
cooling medium. They lie at the lower range of operational parameters for
presently
operating sub- and supercritical fossil power plants (FPP). For these conditions and an
expected component lifetime of 200 000 hours ferritic/martensitic 9-12% Cr steels like
1.4922 and P 91 or HCM12 are used today. For elevated temperatures up to a
maximum of 650°C austenitic stainless steels such as 1.4910, TP 347 HFG and others are
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available. For most of these alloys a broad technical data base and material properties
including creep rupture properties and corrosion behavior are available for subcritial
lifesteam conditions, whereas experience under supercritical water conditions is still
restricted and the data have not yet been published.
A further increase of the operational temperature beyond 650°C would necessitate the use
of highly alloyed Ni-based superalloys with a strongly improved high temperature creep
rupture strength and high oxidation resistance. The development of such high
temperature supercritical fossil power plants is, however, in a premature stage with
unknown results.
It is assumed, as already mentioned in the WPI-Summary report, that in the conventional
part of an HPLWR the well developed and tested materials of the commercial power
plants can be adopted. Therefore, in this report a general overview of these materials and
a short status of knowledge is given. A more detailed evaluation of their oxidation
kinetics was made [4.1] to estimate the corresponding metal loss and the oxide debris to
be expected in the water cycle of the reactor core, where very thin-walled components
and structures like fuel pins and wrappers are exposed to supercritical water. This
evaluation is based on published data from conventional steam power plants. A parabolic
time- and an Arrhenius-type temperature dependence was assumed for the description of
the oxidation behavior in dry steam environment, and for the material loss caused by
spallation of produced oxide scales a linear time dependence was used. Oxide growth
and spallation constants were determined by using fitting algorithms provided by a
standard software for a temperature range between 550°C and 650°C. The derived
formulas allow the calculation of the expected oxide and metal loss and have been
applied to estimate the loss of wall thickness of thin cladding materials during the
expected lifetime of 45 000h. Figure 4-1 gives as an example a comparison of oxidation
and spallation effects at 600°C for selected ferritic/martensitic and austenitic stainless
steels.
A direct application of these findings under steam to the supercritical water conditions in
the conventional part of the HPLWR has to be made with caution. Experience regarding
the general corrosion behavior in the running novel SC FPPs is still limited to low
exposure times and data have not yet been published in the open literature. It is, however,
assumed that like in the subcritical steam regime, also in supercritical water the formation
of stable Cr2O3 layers is the dominant protective mechanism, and the thermodynamic
stability of these protective layers is ensured at least up to 650°C even at high pressure.
This assumption is supported by older data on the corrosion behavior of austenitic
stainless steels in water where the system pressure range has been varied from 70 to 350
bar, and where no substantial differences in corrosion behavior have been observed with
increasing pressure. A direct comparison of older data on the corrosion behavior of
ferritic and austenic steels as well as Ni alloys in degassed supercritical water over a
broad temperature range from 427 to 732°C with those mentioned above, did not lead to
a general agreement, because different temperature and time dependencies were found
and exposure times in the older experiments were fairly short.
Another point that should be considered is the water chemistry to be applied in an
HPLWR. The major restrictions on the specification of water chemistry in this oncethrough cycle come from the necessary limitation of impurities in the feedwater for the
reactor. In the HPLWR core a transition from the sub-to the supercritical state of water
HPLWR – D 13
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occurs, which is connected with a strong decrease of impurity solubility and hence is
responsible for the formation of deposits. Furthermore, it is expected that the radiolytic
water decomposition in an HPLWR will not exceed the values observed in existing
Boiling Water Reactors (BWR´s). Therefore the recommendation is to use the stringent
water specifications existing for BWR´s also for the future HPLWRs. It is necessary,
however, to investigate whether the usual way in conventional power plants to increase
the pH-values of the water by adding ammonia-hydrazines is compatible with the
necessary specifications of the in-core water chemistry.
A general concern is that different forms of stress corrosion cracking (SCC) could be a
problem for the use of steels and especially Ni alloys under high pressure supercritical
water conditions. There exists, however, some proven measures which can reduce this
risk. For austenitic stainless steels which are prone to transgranular stress corrosion
cracking (TGSCC) in high oxygen and chloride containing water it is necessary to obey
the strict limitations of these species through appropriate water chemistry control in the
HPLWR, which is possible. Austenitic stainless steels and especially high Ni-containing
alloys suffer from intergranular stress corrosion cracking (IGSCC). However, by an
appropriate material composition such as an intermediate Ni content, a low carbon
concentration or the use of carbon-binding or “stabilizing” elements like Nb or Ti, the
sensitivity of grain boundaries to IGSCC can be reduced. Nevertheless, one of the most
uncertain areas remains the corrosion behaviour of all materials under supercritical water
conditions and a possible influence on stress corrosion cracking phenomena.
The design data for in-core components compiled in Table 4-1 are very ambitious in
comparison with conventional Light Water Reactors, especially with regard to the high
coolant pressure ( 250 bar) and the increase of the water temperature from 290°C inlet
to 510°C outlet, which causes a transition from the sub- to the supercritical state in the
core. The temperature of the claddings of fuel elements can reach more than 600°C and
the calculated neutron exposure accumulates up to 1.131023 n/cm2 or 60 displacements
per atom (dpa) for an envisaged target of 70 GWd/tU burnup. The high neutron and irradiation associated with this burnup target of the fuel elements leads also to the
formation of undesirable elements like helium and hydrogen via inelastic nuclear
reactions in the alloys .This can, in combination with the displacement of atoms, lead to
changes in the mechanical and micro-structural properties and generate dimensional
distortions in the cladding and core structures. In this respect a HPLWR core resembles
more the operational conditions of a Fast Breeder Reactor (FBR) than a Light Water
Reactor and it was soon very clear that classical Zr-alloys cannot withstand such
conditions.
In a first step a selection of available and promising material groups was made. It is
based on creep-rupture data, corrosion behavior in conventional steam power plants and
on successful application in nuclear reactors. Table 4-2 lists the maximum allowable
temperatures, at a given stress level, for three groups of material which in principle have
the potential to fulfill the requirements as in-core cladding and structural materials in a
future HPLWR. MANET II and P9 1 belong to the well known group of 912%CrMoVNb ferritic/martensitic steels which, as mentioned above, are extensively
and very successfully used in modern steam power plants. Specific alloys of this group
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like types 1.4914, EM 10 and FV 448 have shown an excellent irradiation behavior as
wrapper materials of fuel elements in FBRs up to very high fluence levels. A possible
issue could be an irradiation-induced shift of the ductile-to-brittle transition temperature
(DBTT) into the temperature range of 250-300°C.
The second group of materials comprises of austenitic stainless steels with higher creep
strength and/or improved corrosion resistance. Alloy 1.4970 is a Ti-stabilized 15Cr 15
Ni steel which has been extensively tested in several chemical modifications as cladding
material of fuel elements in European Fast Breeder Reactors up to fluence levels of 150
dpa and Incoloy 800 is a high Cr and Ni containing Ti-stabilized alloy with excellent
corrosion behavior, used very successfully in nuclear steam generators. PE16 and Inconel
718 are typical precipitation-hardened alloys with a high Ni content (Ni-based alloys).
PE 16 has successfully been tested as cladding material in FBR fuel elements. Both
alloys have a high creep strength and show low corrosion in steam environment.
Dependent on the chemical composition and the metallurgical state, they can be prone to
stress corrosion and irradiation-induced high temperature helium embrittlement.
In a further step a comparison of essential material properties was made. Fig. 4-2 gives, as
an example, the ultimate tensile strength for selected alloys as a function of temperature and
in Fig. 4-3 their creep rupture strength for 45 000h endurance is plotted. This leads to a first
estimate which upper temperatures can be achieved by fuel pin claddings made of different
materials for two given stress levels of 100 and 200 MPa respectively. The results are
summarized in Table 4-2 and show that for 100 MPa an upper temperature limit of nearby
600°C is realistic for the best 9-12%Cr steel MANET II, whereas for the austenitic stainless
steels, dependent on their chemical composition, an upper temperature limit ranging from
630-690°C can be achieved. This value is further increased to 720°C by the use of
precipitation hardened Ni-alloys like Inconel 718, whereas PE 16 is comparable with the
better austenitic stainless steels. The temperature limits at a stress level of 200 MPa are also
given in Table 4-2 for comparison. It has, however, to be mentioned that an upper stress
limit exists which is caused by an immediate buckling of thin-walled clad tubes under the
high cooling pressure. This limit has been estimated under somewhat conservative
assumptions to lie for all materials in the range of 150 MPa. This figure also determines the
dimensional design of fuel claddings, especially the minimum wall thickness in dependence
of the pin diameter.
Whereas a first comparison of achievable maximum temperatures for the different alloys
in Table 4-2 is based on a constant maximum compressive stress caused by the cooling
medium, a more detailed analysis of a time-dependent development of stress was to
show, how conservative this estimate is. This time dependence of the exerted stress level
in thin claddings is due to an increase of the inner pressure through fission gas release
which reduces the differential pressure and by a possible increase of the hoop stress
caused by the reduction of the wall thickness through outer corrosion and inner
incompatibility with the fuel material. The calculations performed shows in Fig. 4-4 for a
specific case, that the initial mean hoop stress is decreased with increasing burnup due to
the build-up of the inner pressure of the fuel pin by fission gas release, whereas inner
and outer corrosion retard this tendency. As overall conclusion one can make the
statement that the estimate of maximum achievable temperatures in Table 4-2 are
conservative. The assessment also has taken into account irradiation effects like swelling
and irradiation creep and comes to the conclusion that at a maximum fluence of 60 dpa
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Page 45 of 62
swelling is in the range of about 1-1.5% for all three materials (1.4970, MANET II and
PE 16) and can be tolerated. Less clear is to what extent the relatively high stress level of
and above 100 MPa will reduce the time to rupture properties measured for the unirradiated materials. Very few results, available for austenitic stainless steels, indicate that
under irradiation a reduction of time to rupture has to be expected. Like in the case of
stress corrosion an experimental investigation of this problem is of highest priority in the
next phase of exploration.
The assessment clearly showed finally that for a maximum temperature of 650°C from a
standpoint of creep-rupture strength, corrosion resistance and irradiation behavior, not
only Ni-alloys, but also austenitic stainless steels like alloy 1.4970 can fulfill the
requirements for in-core cladding and structural materials. Taking into account specific
items like the neutron absorption, the sensitivity to irradiation-induced helium
embrittlement and stress corrosion cracking, it was finally concluded that the austenitic
stainless steels are the better choice.
The assessment has finally shown that the most uncertain areas in the present analysis are
the corrosion behavior under supercritical water conditions, including the effects of water
chemistry/radiolysis, and the influence of a high stress state on stress corrosion and
deformation mechanisms which govern the creep-rupture and creep buckling properties.
These activities should in a later stage be expanded to in-reactor experiments in order to
investigate the effect of irradiation. For a further development of in-core materials the
proposed class of recommended austenitic stainless steels has to be further optimized with
regard to type and degree of stabilization, the necessary carbon content and the balance
between major and minor alloying elements to achieve an optimum in creep rupture
strength, corrosion resistance and stability under irradiation. For the long term the further
development of dispersion-strengthened ferritic steels is promising.
All these activities should be part of a Key Technology Phase to be started in the next
European Framework Programme of the HPLWR Project.
Table 4-1:
HPLWR ”Reference Design” Data for in-core, RPV, and
ex-core components
In core data
Coolant
Coolant pressure [MPa]
Coolant inlet/outlet temperature [°C]
Fuel/Enrichment
Fuel/Enrichment, revised [5]
25
280/508
UO2/5%
MOX/ to be determined
Burnup [GWd/tU]/ lifetime [hours]
Burnup [GWd/tU]/ lifetime, revised [5, 6]
45/30,000
70/45,000
Neutron flux [n/cm2s] /fluence [n/cm2]
Cladding outer-diameter/thickness [mm] [2]
Cladding max. surface temperature [°C] [2]
51014/ 81022
8/0.4
620 for Ni alloys
450 for stainless steels
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 4 [2]; Revised [5, 6]:  8
Pin pre-pressurization [MPa]
Potential core structure and cladding materials
Austenitic stainless steels
Ferritic/martensitic steels
Ni-based alloys
1.4550, 316L(N), 1.4970
1.4914, FV 448, EM10
PE 16, Inconel 625, Inconel 718
Reactor pressure vessel
Coolant pressure [MPa]
Temperature [°C]
25-27.5
350
Lifetime [years]
Materials
60
Ferritic steels ( 20 MnMoNi 5 5)
Ex-core data
Life steam pressure [MPa]
Life steam/reheat temperature [°C]
25-27.5
540/560
Lifetime [h]
Materials
Ferritic/martensitic steels
200,000
X20 CrMoV12 1, P91, E911, P92
(NF616), P122 (HCM12A)
1.4910, TP347HFG, Super304, NF709,
Incoloy 800 HAT
Austenitic stainless steels
Table 4-2:
Estimated maximum temperatures for different materials for the
condition of RM/45,000 h at 100 MPa and 200 MPa respectively
Stress+
Temperature
Reference
MANET II
MANET II
EURALLOY
EURALLOY
1.4970
15Cr-15Ni-Ti
(sa + cw + a)
Incoloy 800
100
200
100*
200*
100
200
587
512
553
494
690
629
FZKA 5722
1996
AGT1-SG2-03
1992
KfK 4217
1986
100
200
625
544
Inconel 718
100
200
100
200
712
672
690
650
MM Werkstoffblatt 760
1976
PSB 354
1970
AGT1-SG2-1
1992
Material
PE 16
+
Without any safety margin!
* For equivalent stress level  (von Mises)
HPLWR – D 13
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HPLWR Contract No. FIKI-CT-2000-00033
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Oxidation & spallation of steels at 600ºC
OXIDE LAYER THICKNESS, d (µm)
90
HCM12A
80
1.4910
70
TP347HFG
60
50
40
30
20
10
0
0
5000
10000 15000 20000 25000 30000 35000 40000 45000 50000
TIME (h)
Figure 4- 1
A comparison of oxidation and spallation for ferritic and austenitic steels
at 600°C. Metal loss is half of the oxide thickness.
900
850
800
750
Ultimate tensile strength (MPa)
700
650
600
550
500
450
400
350
300
250
200
Manet II
150
DIN1.4970
100
PE 16
Inconel 625
50
0
0
100
200
300
400
500
600
700
800
Temperature (°C)
Figure 4- 2
Ultimate tensile strength RM for selected alloys as a function of temperature
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600
1.4970
Creep rupture strength Rm/45,000h (MPa)
550
MANET-II
Incoloy 800
500
Inconel 718
450
400
350
300
250
200
150
100
50
0
400
450
500
550
600
650
700
750
800
Temperature (°C)
Figure 4- 3
Creep-rupture strength RM/45,000h for selected alloys
Mean hoop stress [MPa]
sme = 0.4mm outer corr.
-50 (1)
(3) sme = 0.7mm outer corr.
(2) sme = 0.4mm outer+inner corr.
(4) sme = 0.7mm outer+inner corr.
(3)
-100
(4)
-150
(1)
(2)
HCM12A, T=650°C, po=25 MPa
OXSPA
-200
0
Figure 4- 4
10,000
20,000
30,000
Time [h]
40,000
50,000
The evolution of the mean hoop stress in the cladding by comparing the
effect of outer corrosion with the combined outer and inner corrosion
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4.4
References
4.1
Ehrlich, K.,Konys,J.,Heikinheimo,L.,Arnoux, P.,Schirra, M., Leistikow, S.
Steiner, H., “Preliminary Evaluation of Candidate Materials Against the
Combined Effects of Creep and Corrosion: In-core and Out-of-core Materials
Selection for the HPLWR”, HPLWR-D8, August 2002 (RE)
Leistikow, S, Konys, J., Ehrlich, K., “Available Metallurgical and Process
Parameters/Treatments for Reduction of General Corrosion/Stress Corrosion
Cracking for In- and Ex-vessel Application”, HPLWR-D 9/10, August 2002 (PU)
4.2
HPLWR – D 13
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Page 50 of 62
Work Package V – Economics
D. Bittermann(Framatome ANP), D. Squarer(FZK), A. Souyri(EdF), Y. Oka(U of Tokyo)
5.1
5.2
-
Objectives
To identify and analyze the most important parameters which influence plant and fuel
cycle cost
To estimate the economic potential of a HPLWR
Description of Work
Study and evaluate fuel cycle costs of a HPLWR and identify specific differences to
LWR concepts
Identify the main components, systems and equipment which may not be needed
compared to BWR and PWR plants and simplify design
Review the economic study of the supercritical water reactor done in Japan and
estimate the HPLWR economic target (for given boundary conditions) which could
be reached
5.3
Deliverables and Milestones
HPLWR-D11 [5.2]
5.4
Review of other economic studies
5.4.1 SCLWR economic study
An economic evaluation performed for the Oka concept [5.1] has been reviewed. In this
evaluation, the SCLWR was compared with the ABWR. It was assumed that the material
selection and design issues are solvable with appropriate investment and that design
issues such as the water rods are solved. It was also assumed that the cost of the reactor
building varies according to the base area of the building (36% reduction for the
SCLWR) and that similar construction time is required for both the SCLWR and the
ABWR. In comparing a 1570 MWe SCLWR plant with a 1350 ABWR plant it was found
that a saving will occur due to smaller reactor building (due to smaller reactor vessel),
smaller cooling tower (1/3 smaller heat rejection), and less LP turbine and condenser
(smaller mass flow). The turbine building and plant layout are similar for both plants but
more BOP equipment is needed for the SCLWR due to startup conditions.
The construction cost of SCLWR is comparable to ABWR and when normalized by the
power output is 15% lower than ABWR. The following major conclusions were drawn in
Japan’s SCLWR economic study: major cost reduction is due to smaller reactor building;
first of a kind plant will cost more initially; production cost (O&M + fuel cycle) is
expected to be higher than the ABWR due to higher O&M cost and 8 % higher fuel cycle
cost; owners cost are to be determined; the cost of equipment in the turbine building is
about 20 % of the total plant cost; the SCLWR turbine cost and feed-water system
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components yielded 30 % higher cost; the SCLWR cost resulted in 1570 $/kWe vs. 1850
$/kWe for the ABWR or about 15% reduction; the total normalized (by power) direct
cost were: 1010 $/kWe for the SCLWR vs. 1210 $/kWe for the ABWR, i.e. a reduction
of 17%.
5.4.2 Comparison with other existing concepts
Existing designs of advanced LWR are simplified, more economical (less than
$1500/kWe, 3-4.5¢/kWh), safe, certified design (pre-licensed), shorter construction time,
temperature and pressure are lower than the HPLWR [5.2].
The generation cost of operating large scale power plants (1999 US data) are:
–
Nuclear 1.83, Coal 2.07, Oil 3.18, Gas 3.52 ¢/kWh
–
Nuclear fuel only 0.481 ¢/kWh (US average)
–
Fuel average cost: nuclear 0.5, coal 1.45, oil 2.41, gas 2.84 ¢/kWh
The projected cost of the proposed HTGR (pebble-bed) is ~$1100/kWe for a 110 MWe
PBMR with 41% efficiency and 1.5- 3 years construction time (modular).
5.4.3 GE’s Advanced BWR for improved economics
The General Electric Corporation has improved continuously the economics of its
advanced BWR [5.4]. It is interesting to examine the main characteristics of these
improvements in order to help focus the HPLWR economics.
Factors resulting in improved economics of the advanced BWR that meets the EUR are:

Simpler structures, higher margins, easier construction
•
Higher power density, higher plant power, use of modular passive safety systems
•
Design features enhancing economy of scale: Gravity Driven Cooling System
(GDCS) pool as a part of the wetwell, modular safety systems with little dependence on
power level, smaller Passive Containment Cooling System (PCCS) pools and larger heat
exchangers
•
Improved design: large blade control rods, simpler reactor internals, improved
plant arrangements (moved non-safety systems, stacked spent fuel, flexible building
embedment-external cask hatch)
•
Simplification studies, while maintaining performance margins:
– Reduced fuel bundles, Control Rod Drive (CRD), vessel, increase fuel
length, significant simplification in vessel and internals
– Improved plant availability (5%)- refueling and outage plan and system
improvements
– Reduced building and structures (30%)- reduce basemat thickness, reduce
containment design pressure, move spent fuel pool to grade
elevation/separate building, separate reactor building from containment
•
Building and refueling optimization:
– Building size is controlled by: wetwell, PCCS parameters, Main Steam
Isolation Valve (MSIV) access control
– Vessel height does not control building height
– Refueling floor size and dimensions control footprint
– Refueling schemes are important for optimization
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–
–
–
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The structures are controlled by containment design pressure and plant
seismic design basis. Find ways to reduce design pressure
Reduce construction schedule
Spent fuel in containment or reactor building: horizontal or inclined fuel
transfer, stacked spent fuel option, cask transfer schemes, size of spent
fuel pool, refueling floor arrangement, location of steam line
DOE’s Near Term Deployment Economics
The US DOE has examined recently (October 2001) the economics of advanced LWR
plant that can be deployed by 2010, in comparison with Combined Cycle Gas Turbine
plant (CCGT) and with Gas Turbine plant (GT) [5.2]. The following results were
obtained:
•
Life cycle generation cost of 3.6-4.6 ¢/kWh is competitive with market
prices and with combined cycle gas fired plants. Production cost of 0.5 ¢/kWh for fuel
cycle and 0.5 ¢/kWh for Operation and Maintenance (O&M) are projected for Advanced
LWR (ALWR), i.e. production cost of 1 ¢/kWh is ~24% of life cycle generation cost.
The rest is “cost of money” (Return On Investment (ROI), etc.)
•
Advanced nuclear power plants can compete with all types of fossil power
plants in deregulated markets
•
Costs in the early years of life should be resolved by longer term
commitments to purchase power
The results of the levelized generation cost for the three types of power plants are shown
in the following Table 5.1 and in Figure 5.1.
Table 5.1 Cost comparison of ALWR and gas turbine plants
ALWR
CCGT $600/kWe GT $300/kWe
1000$/kWe EPC
$4/MMBTU
Total cost ¢/Kwh 4.49
4.58
6.42
Capital cost
¢/Kwh
3.49
1.58
2.32
IRR (part of
capital cost)
1.78
0.49
0.35
O&M cost
0.5
0.2
0.1
Fuel cost
0.5
2.8
4.0
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Levelized Generation Cost in
cents/kwh (IRR=Internal Rate of
Return)
Generation cost in
cents/kwh
7
6
Total cost ¢/kWh
5
Capital cost
4
IRR cost
O&M cost
3
2
Fuel cost
1
0
ALWR
CCGT
GT
Power Plant type
Cost comparison of Advanced LWR (ALWR) with Combined Cycle
Gas Turbine (CCGT) and Gas Turbine (GT) (DOE Near-Term Deployment,
October 2001)
Figure 5. 1
In order for a new near-term deployment nuclear power plants to be economical the
following conditions are suggested:
•
4 years lead construction time, 5 years total project lead time
•
Resolution of licensing issues before project commitment
•
Total overnight capital cost, including owner cost and contingency, of 1,1001,500 $/kWe (depending on assumed fossil fuel prices)
•
Typical 1000 MWe nuclear plant requires a total as-spent investment (current
year $) of ~$2x109 ($2 B), hence the competition from low up-front investment gas fired
plants
•
Nuclear plants have low and stable running costs that are ideal for long-term bilateral power purchase contracts
•
Nuclear plant life-time capacity factor of 85-90% and long plant operating life
(40-60 years)
•
High safety performance also assures better economics
•
Plant site impacts the economics. Select the location where market prices would
exceed 4 ¢/kWh
•
Nuclear plant economics is strongly impacted by the financial package. Minimize
%ROI, extend debt repayment, reduce equity financing to 40% or lower
HPLWR – D 13
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In addition, in order to minimize the “time to market”, it is desirable to have:

Efficient regulatory approvals

Parallel regulatory approvals and design completion

Early procurement of long-lead components
Furthermore, Government incentive to reduce business risk could follow along the
following lines:

Encourage long-term power purchase agreements

Accelerated depreciation

Tax credit for new investments

Tax incentive for fuel supply diversity and emission-free generation

Access to tax-exempt government financing

Ensure energy/environmental policies and regulations are balanced
5.5 Considerations concerning the economics of a HPLWR plant
5.5.1 General
Japan’s SCLWR economic study made several key assumptions that should be
substantiated before drawing final conclusions. It considered a specific design [5.1] in
considerable details. Thus any changes in plant configuration, plant systems, plant size
(i.e. electrical power output), containment, fuel assembly design, fuel enrichment and fuel
composition, in-vessel and ex-vessel materials, safety systems, mitigating severe accident
features, passive safety systems, etc. would impact the economic feasibility. Such
changes are foreseen due to numerous impractical features of the SCLWR in a European
arena. It is therefore necessary to define the HPLWR plant in more details before
attempting to evaluate its economic feasibility. Since many of the above listed items are
expected to differ between the HPLWR and the SCLWR, the economic feasibility of the
HPLWR remains an open question.
Several major plant systems and equipment that may not be needed in a once-through
SCLWR compared to PWR and BWR were identified by Prof. Oka in several
publications [5.1] (e.g. steam generators, recirculation pumps, steam separators,
pressurizer). However, the above list of variances from the SCLWR could add additional
systems. Since the ABWR already includes many of these features in accordance with the
European Utilities Requirement document (EUR), and the economic feasibility is
determined in comparison with the ABWR, these additional features may not by
themselves make the HPLWR uneconomical.
A significant economical benefit can be obtained when ‘off-the-shelve’ equipment is
used. Thus, if the plant is being designed with this in mind a significant development cost
can be avoided. This may be true for example, for the turbine design, the reactor pressure
vessel, valves, chemical plant, etc.
It should be mentioned here that a substantial effort and investment (hundreds of million
of Euro) were made during the last decade in order to simplify and make the ABWR
more economical. This program culminated in the design of the SBWR and the ESBWR.
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Thus, some of the methodologies that were applied to the economical design of the
ESBWR may also be applied to the design of an HPLWR.
An analysis for evaluation of the electricity generation costs has been performed and the
findings of the economic evaluation [5.2] are described below.
5.5.2 Fuel Cycle Considerations
The evaluation and final definition of the fuel type, fuel enrichment, burnup and fuel
assembly design have not been completed. However, a significant progress was made
towards the definition of these parameters [5.3]. Therefore a rough analysis of the fuel
cycle cost was performed. As a reference the fuel cycle cost of a PWR is assumed for
which the cost structure and specific costs are well-known. For the HPLWR currently it
is considered that there are three major cost elements which have an essential
contribution, yet are not exactly known. These are the Fuel Assembly (FA) fabrication
cost, the enrichment and the efficiency. Therefore these cost contributors were assumed
as parameters and a variation within reasonable ranges was performed. The conclusions
of this analysis are as follows [5.4]:
 For the targeted burn-up of 60 GWd/kgU, an assumed efficiency of 44% and FA
fabrication costs of 350$/kgU which are considered as reasonable, fuel cycle cost (FCC)
of a HPLWR is below that of PWR FCC up to average enrichments of 6.5%
 Both significant influence of efficiency and FA fabrication costs on FCC has to be
considered
 Significant influence of uranium ore price exists only if price increases by factors;
relative to PWR the difference in FCC becomes larger with price increase due to the
effect of higher efficiency of the HPLWR
 Design measures which reduce enrichment at the expense of reduction of the
efficiency is expected to be more beneficial than keeping the efficiency high
In addition the cost of a MOX fuel cycle were analyzed. The fuel fabrication was
assumed at 2500 $/Kg and the disposal cost at 250 $/Kg (reprocessing cost was not
included). The calculation indicated that the use of Pu will cost more than using enriched
uranium. The cost varied linearly with the enrichment starting approximately at 2.5
mills/kWh for 5% and reaching ~4.7 mills/kWh for 10%.
5.5.3 Specific evaluation of HPLWR electricity generation costs
The major contributors to Nuclear Power Plant (NPP) costs are [5.2, 5.5]: capital cost
about 60-70%, O&M cost about 17-25% and fuel cost about 13-15%. Generally the
capital cost of NPP is made of: Direct cost (e.g. reactor plant equipment, etc.), Indirect
cost (e.g. design and engineering, etc.), Other cost (e.g. training, taxes and insurance,
etc.). The total capital cost is the sum of these three categories and is equal to the
overnight cost since the financial cost is not included. For a typical PWR the breakdown
of these costs is: Direct cost- about 75%, Indirect cost-about 14% and Other costs- about
11%. Cost breakdown was estimated for the HPLWR based on the published results for
the ABWR and a PWR. The estimated costs breakdown is: Direct cost- 70% for the plant
which was used as reference for the HPLWR (HPLWR Reference), 75% for a PWR and
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63% for the ABWR; Indirect cost- 23% for the HPLWR Reference, 14% for a PWR and
22% for the ABWR; Other cost- 7% for the HPLWR Reference, 11% for a PWR and
15% for the ABWR. Major cost reductions are expected due to: no- steam separators,
steam dryers, circulation pumps; smaller-RPV, containment, reactor building, number of
steam lines, spent fuel pool volume, condenser, cooling tower, building volume, I&C
cost, crane capacity, capacity of HVAC, fire protection measures. Major costs increases
are expected due to: equipment for start-up, design of specific components and systems
for higher pressure and temperature, steam reheater with extended design conditions,
complicated fuel assembly. The cost impact of the technical changes was evaluated by
means of published scaling factors for individual components. Based on this analysis the
estimated cost saving for a1000 MWe HPLWR due to the above mentioned factors, is
approximately 23%, with the following estimated cost reduction: 25% in capital cost,
25% in O&M cost, 10% in fuel cost. The following breakdown was estimated for the
HPLWR fuel cycle cost (3.08 mills/kWh): 39.2% for enrichment, 25.6% for disposal,
0.3% for conversion, 12.9% for fabrication and 22.1% for uranium ore. The economic
evaluation [5.2, 5.5] also included calculations showing at what specific capital cost a
nuclear power plant becomes competitive with a coal power plant or a combined cycle
plant. For a coal fired plant with a coal price of 40 €/t, NPP will be competitive at
specific capital cost lower than ~1500 €/kWe, whereas for combined cycle plants of
medium and large size with gas price of 0.1€/m3 and an interest rate of 5%, NPP is
competitive when the specific capital cost is ~1500 €/kWe (~1100 €/kWe at 10%
interest). If the gas price falls below about 0.07€/m3, NPP can not compete with medium
and large size combined cycle plants, however NPP can compete with coal fired plants
even at a coal price of 30 €/t. For small size combined cycle plants, NPP can compete
even at a gas price of 0.05 €/m3 if an NPP specific capital cost can be kept at ~1000
€/kWe. In conclusion, based on the current status of the HPLWR, the capital cost of a
1000 MWe plant has the potential to be 20-25% lower than advanced LWRs. The cost of
electricity generation by a HPLWR is considered to have an advantage compared to all
fossil power plants, and considering the potential future increase in fossil fuel costs the
advantages of the HPLWR generation cost can be significant.
5.4
Conclusions
Cost reduction is the most essential requirement for nuclear power plants in order that
their application would expand in the future. Currently only very large nuclear plant sizes
are economical but they do not have big advantages in electricity generation costs
compared to that of fossil plants with nowadays fuel prices. Therefore the vendors of
nuclear plants are looking for concepts with the potential for significant economic
advantages compared to fossil fired plants. This is especially true for plant sizes less than
1000 MWe.
The estimated cost reductions for the HPLWR compared with the defined reference plant
are: 30% reduction for building and structures, 35% reduction for the reactor plant, 10%
reduction for the turbine plant, and 20 to 25% reduction in overnight capital cost. An
initial economic target for the HPLWR is set at 1000 €/kWe and 3-4 cent/kWh levelized
generation cost.
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The economic evaluation of the HPLWR plant concept presented in this report, indicates
that such nuclear plants may have the potential to reach electricity generation costs which
can significantly increase the economic advantages of nuclear plants compared with
fossil fueled plants. This statement especially holds true if one considers in the future a
further increase in fossil fuel prices and the possibility of getting deeper insight into the
design of the HPLWR in order to explore the potential of cost reduction more precisely
and in more detail.
Considering the current technical status of the HPLWR, this evaluation is considered to
be substantiated enough to justify a proposal for the continuation of the next step of
HPLWR development work.
5.5
5.1
5.2
5.3
5.4
5.5
References
Dobashi, K., Oka, Y., Koshizuka, S., “Conceptual Design Of A high Temperature
Power Reactor Cooled And Moderated By Supercritical Light Water”, ICONE 6,
May 1998, ASME, NY, NY
Bittermann, D. (Framatome ANP, Erlangen, Germany), Squarer, D.
(Forschungszentrum Karlsruhe, Germany), “Preliminary Economic Evaluation of
The HPLWR“, HPLWR-D11, July 2002 (RE)
Rimpault, G. et. al., “Results of Evaluation of Existing Core Designs and Potential
for Applications for HPLWR”, HPLWR-D4, September 2002 (RE)
Squarer, D. “Minutes of The Third HPLWR Project Meeting of August 27-30,
2001 at PSI – Würenlingen and Villigen, Switzerland”, Forschungszentrum
Karlsruhe, Germany, HPLWR-M08, September 2001 (RE)
Squarer, D., “Minutes of The Fifth HPLWR Project Meeting of July 29-31, 2002
at FZK – Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M 11, September 2002 (RE)
HPLWR – D 13
6.
HPLWR Contract No. FIKI-CT-2000-00033
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Conclusions
D.Squarer(FZK), D. Bittermann(Framatome ANP), Y. Oka(U of Tokyo), P. Dumaz(CEA),
G. Rimpault(CEA), R. Kyrki-Rajamaki(VTT), K. Ehrlich(FZK-MCS), N. Aksan(PSI), C.
Maraczy(KFKI), A. Souyri(EdF)
At the conclusions of the HPLWR project, it is clear that all the major objectives set up at
the begining of the project were achieved. The effort invested by the partners in the
HPLWR project far exceeded the effort originally foreseen. The list of publications and
reports generated by the HPLWR project (see Appendix A) is a testimony to the
substantial and productive output of the project.
The major conclusion of the project is that the HPLWR concept has a technical merit and
a potential to be economically feasible in comparison with other nuclear or fossil power
plants. An initial (believed to be achievable) economic target for the HPLWR is set at
1000 €/kWe and 3-4 cent/kWh levelized generation cost.
Despite the substantial technical progress made by the HPLWR project, a lot more
remains to be done in the future before it can be introduced to the market place by
vendors. It is hoped that future support for the HPLWR will continue under the
Commission’s 6th FP, thus bringing the concept closer to the market place. This is
particularly important since the HPLWR concept is receiving recently significant
attention and support in the USA and in Japan.
The HPLWR project has identified the following potential future activities:
Key technology program, accompanied by a preliminary design under the 6th FP, to be
followed by a detailed design of the HPLWR.
The following key technologies were highlighted by the project: cladding material for
temperatures up to 650 ºC; enhanced heat transfer in the fuel assembly under supercritical
water conditions in order to reduce the maximum cladding temperature; thermalhydraulics of the core and fuel assembly; improved coupled neutronics/thermalhydraulics codes for fuel assembly, core and plant analysis; development of a simplified
sub-assembly design including consideration of extended burn-up on reactivity
coefficients; optimization of the plant safety systems, maximizing the passive safety
features and verifying the computer codes necessary for safety analysis; neutron physics
experiments for computer codes and fuel assembly design verification; evaluation of the
HPLWR as a fast reactor; cost performance and economic optimization of the HPLWR;
training of young scientists and engineers in key technology areas.
The following key technology experiments were identified by the HPLWR project:
Creep tests of selected cladding materials; corrosion and stress-corrosion cracking tests
for selected cladding materials under supercritical steam conditions and the HPLWR
specific water chemistry; heat transfer tests of the fuel assembly at supercritical water
conditions under steady-state and transient conditions; critical flow tests for pipe break
under supercritical water conditions; verification tests of containment design in order to
HPLWR – D 13
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verify containment analysis codes; reactor physics verification experiments (e.g. in
PROTEUS, EOLE, ILL facilities).
The following is a brief summary of recommended HPLWR future activities:
HPLWR concept refinement and assessment; HPLWR at higher neutron energies (fast);
accurate and extensive core design effort; benchmark and validation of computer codes,
including experiments in: neutron physics, sub-channel thermal-hydraulics, deteriorated
heat transfer, transient, safety and corrosion; consideration of European requirements and
guidelines for future reactors (for example EUR, Technical Guidelines) for the HPLWR
in particular with respect to safety criteria; iteration on the HPLWR design to reduce cost,
including fuel cycle cost and multi-purpose concepts.
HPLWR – D 13
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Appendix A : List of HPLWR Reports, Minutes and Memos
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
Squarer, D., “Minutes of the Kick-off Meeting of August 29-30, 2000”, Forschungszentrum
Karlsruhe, Germany, HPLWR-M01, September 2000 (RE)
Oka, Y., “Review of High Temperature Water and Steam Cooled Reactor Concepts”,
Nuclear Engineering Research Laboratory, The University of Tokyo, Japan, HPLWR SR01, October, 2000 (PU) (Also pp. 37-57 of the Proceedings of the First International
Symposium on Supercritical Water-cooled Reactors, Design and Technology, SCR-2000,
The University of Tokyo, Tokyo, Japan, November, 2000)
Heusener, G. Muller, U., Schulenberg, T., Squarer, D. “A European Development
Program for a High Performance Light Water Reactor (HPLWR)”, SCR-2000, The
University of Tokyo, Tokyo, Japan, November 2000
Squarer, D., “Reasons For Selecting The University Of Tokyo Reactor Design As A Study
Reference”, Forschungszentrum Karlsruhe, Germany, HPLWR-MEM01, December, 2000
(RE)
Bittermann, D., Squarer, D., “Minutes of the WP I Working Group Meeting of January 1819, 2001 At Forschungszentrum Karlsruhe, Germany”, Forschungszentrum Karlsruhe,
Germany, HPLWR-M 02, February, 2001 (RE)
Aksan, N., Schulenberg, T., Cheng, X., “Minutes of the WP-III Technical Meeting of
January 29-30, 2001, At Forschungszentrum Karlsruhe, Germany”, Paul Scherrer Institut,
Switzerland and Forschungszentrum Karlsruhe, Germany, HPLWR-M 03, February, 2001
Squarer, D. “Minutes of The Second HPLWR Project Meeting of March 5-6, 2001 (Rev.
1) at CEA – Cadarache, France”, Forschungszentrum Karlsruhe, Germany, HPLWRM04, April 2001 (RE)
Squarer, D., “HPLWR Management Report No. 1”, HPLWR-P1, March 1, 2001(RE)
Schulenberg, T., “Minutes of the WP I Meeting of April 27, 2001 At Forschungszentrum
Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany, HPLWR-M 05, May,
2001 (RE)
Konys, J., Ehrlich, K., “Minutes of the WP IV Working Group Meeting of April 11, 2001 At
Framatome ANP, Erlangen, Germany”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M 06, May, 2001 (RE)
Cheng, X., Schulenberg, T., “Heat Transfer At Supercritical Pressures-Literature Review
and Application to an HPLWR”, Forschungszentrum Karlsruhe, Technik und Umwelt,
FZKA 6609, May 2001
Dumaz, P., “Contribution to the Analysis of Core Design Constraints in SupercriticalPressure Light Water Reactors”, CEA Technical Note NT-SERI-LFEA-01, May, 2001,
HPLWR-D15 (RE)
Rimpault, G., Testa E., “Neutronic contribution to the Oka reference design evaluation”,
CEA/Cadarache Center, France, May 2001, HPLWR-D16 (RE)
Rimpault, G. “Experimental Programmes For Assessing HPLWR Neutronic Calculations”,
CEA/Cadarache Center, France, May 2001(RE)
Konys, J., Ehrlich, K., “Minutes of the WP IV Working Group Meeting of June 19, 2001 At
Forschungszentrum Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M 07, July, 2001 (RE)
X. Cheng, T. Schulenberg, A. Souyri, V. Sanchez, “Heat Transfer and Pressure Drop in
Supercritical Pressure- Literature Review and Application to a HPLWR”,
Forschungszentrum Karlsruhe, Germany and EdF, France, HPLWR-D 07, August, 2001
(PU)
D. Bittermann, “General Plant Characteristics- Draft”, Framatome ANP, Erlangen,
Germany, HPLWR-D 03, August, 2001 (RE)
D. Squarer (FZK, Karlsruhe), Y. Oka (Univ. of Tokyo), D. Bittermann (Framatome ANP,
Erlangen), N. Aksan (PSI, Villigen), C. Maraczy (KFKI, Budapest), R. Kyrki-Rajamaki
(VTT, Espoo), A. Souyri (EDF, Paris), P. Dumaz (CEA, Cadarache), “HIGH PERFORMANCE
LIGHT W ATER REACTOR (HPLWR)”, FISA-2001, November, 2001(PU)
HPLWR – D 13
19.
20.
21.
22.
23.
24.
25.
26.
27.
28.
29.
30.
31.
32.
33.
34.
35.
36.
37.
HPLWR Contract No. FIKI-CT-2000-00033
Page 61 of 62
Tanskanen, A., Wasastjerna, F., “VTT’s Contribution to the HPLWR Neutronics Studies”,
HPLWR-D17, August 2001 (RE)
Yamaji, A., Oka, Y., Koshizuka, S., “Conceptual Core Design of a 1000 MWe
Supercritical-Pressure Light Water Cooled and Moderated Reactor”, Paper presented at
the ANS and HPS meeting on March 29-April 1, 2001, at Texas A&M University
Technological Implementation Plan (TIP)-Preliminary Version At Mid-term, August, 2001
Squarer, D., “HPLWR Management Report No. 2”, HPLWR-P2, September, 2001 (RE)
Squarer, D. “Minutes of The Third HPLWR Project Meeting of August 27-30, 2001 at PSI
– Würenlingen and Villigen, Switzerland”, Forschungszentrum Karlsruhe, Germany,
HPLWR-M08, September 2001 (RE)
D. Squarer (FZK), D. Bittermann (Framatome ANP), Y. Oka (U. of Tokyo), P. Dumaz
(CEA), G. Rimpault (CEA), R. Kyrki-Rajamaki (VTT), K. Ehrlich (FZK-MCS), N. Aksan
(PSI), C. Maraczy (KFKI), A. Souyri (EdF), “HPLWR Annual Technical Report”, HPLWRD12, September 2001 (RE)
D. Squarer (FZK, Karlsruhe), Y. Oka (Univ. of Tokyo), D. Bittermann (Framatome ANP,
Erlangen), N. Aksan (PSI, Villigen), C. Maraczy (KFKI, Budapest), R. Kyrki-Rajamaki
(VTT, Espoo), A. Souyri (EDF, Paris), P. Dumaz (CEA, Cadarache), “High Performance
Light Water Reactor (HPLWR)”, FISA-2001, Luxembourg, November 12-15, 2001
N. Aksan (Paul Scherrer Institut, Villingen), D. Bittermann (Framatome ANP, Erlangen),
P. Dumaz (CEA, Cadarache), R. Kyrki-Rajamaki (VTT, Espoo), C. Maraczy (KFKI,
Budapest), Y. Oka (University of Tokyo), T. Schulenberg, D. Squarer
(Forschungszentrum Karlsruhe), A. Souyri (EDF, Paris), “A High Performance Light
Water Reactor Concept ”, Annual Meeting on Nuclear Technology, Stuttgart, May, 2002
Konys, J., Ehrlich, K., “Minutes of the WP IV Working Group Meeting of November 29,
2001 At Forschungszentrum Karlsruhe, Germany”, Forschungszentrum Karlsruhe,
Germany, HPLWR-M 09, December, 2001 (RE)
X. Cheng, T. Schulenberg, S. Koshizuka, Y. Oka, A. Souyri, „Thermal-Hydraulic Analysis
of Supercritical Pressure Light Water Reactors”, International Congress on Advanced
Nuclear Power Plants (ICAPP), American Nuclear Society, Florida, June 9-13, 2002
K. Ehrlich (FZK, Karlsruhe), L. Heikinheimo (VTT, Espoo), P. Arnoux (CEA, Saclay), D.
Bittermann (Framatome ANP, Erlangen), Y. Oka (University of Tokyo), T. Schulenberg
(FZK, Karlsruhe), “Material Requirements of High Performance Light Water Reactors”, 4 th
Workshop on LWR Coolant Water Radiolysis and Electrochemistry, Avignon, France,
April 26, 2002
T. Nakatsuka, Y. Oka, S. Koshizuka, “Startup Thermal Considerations For SupercriticalPressure Light Water-Cooled Reactors”, Nuclear Technology, Vol. 134, pp. 221-230,
June 2001
Squarer, D., “HPLWR Management Report No. 3”, HPLWR-P3, March, 2002 (RE)
Squarer, D. “Minutes of The Fourth HPLWR Project Meeting of March 4-6, 2002 at EdF –
Chatou, Paris, France”, Forschungszentrum Karlsruhe, Germany, HPLWR-M 10, April
2002 (RE)
Oka, Y., Koshizuka, S., “D1: Report on Current Status of Work Concerning Reactor
Designs under Supercritical-Water Conditions”, Nuclear Engineering Research
Laboratory, The University of Tokyo, Japan, HPLWR- D1, April, 2002 (PU)
K. Ehrlich, L. Heikinheimo, M. Schirra, S. Leistikow, J. Konys, P. Arnoux, „HPLWR
Annual Technical Report 2001, Work Package IV- Material and Corrosion“, FZK Interner
Bericht, 32.22.09, NUKLEAR 3363, October 2001
D. Bittermann, “Current and Future Turbine Technology Under Supercritical Water
Conditions”, Framatome ANP, Erlangen, Germany, HPLWR-D 02, August, 2002 (RE)
D. Bittermann (Framatome ANP, Erlangen, Germany), D. Squarer (Forschungszentrum
Karlsruhe, Germany), “Preliminary Economic Evaluation of The HPLWR“, HPLWR-D11,
July 2002 (RE)
K. Ehrlich(MCS, Karlsruhe, Germany), J. Konys(FZK, Karlsruhe, Germany), L.
Heikinheimo(VTT, Espoo, Finland), P. Arnoux(CEA, Saclay, France), M. Schirra (FZK,
Karlsruhe, Germany), S. Leistikow (Karlsruhe, Germany), H. Steiner(FZK, Karlsruhe,
Germany), “Preliminary Evaluation of Candidate Materials Against the Combined Effects
HPLWR – D 13
38.
39.
40.
41.
42.
43.
44.
45.
46.
47.
48.
49.
50.
HPLWR Contract No. FIKI-CT-2000-00033
Page 62 of 62
of Creep and Corrosion: In-core and Out-of-core Materials Selection for the HPLWR”,
HPLWR-D8, August 2002 (RE)
G. Rimpault(CEA, Cadarache, France), P. Dumaz(CEA, Cadarache, France), C.
Maraczy(KFKI, Budapest, Hungary), A. Tanskanen(VTT, Espoo, Finland), F.
Wasastjerna(VTT, Espoo, Finland), R. Kyrki-Rajamaki(VTT, Espoo, Finland), Y. Oka(U.
of Tokyo, Tokyo, Japan), S. Koshizuka (U. of Tokyo, Tokyo, Japan), C. Broeders(FZK,
Karlsruhe, Germany), A. Bergeron(CEA, Saclay, France), E. Kiefhaber(FZK, Karlsruhe,
Germany), D. Sruwe(FZK, Karlsruhe, Germany), P. Rau(Framatome ANP, Erlangen,
Germany), X. Cheng(FZK, Karlsruhe, Germany), T. Schulenberg(FZK, Karlsruhe,
Germany), V. Sanchez(FZK, Karlsruhe, Germany) “Results of Evaluation of Existing
Core Designs and Potential for Applications for HPLWR”, HPLWR-D4, September 2002
(RE)
G. Rimpault(CEA, Cadarache, France), P. Dumaz(CEA, Cadarache, France), C.
Maraczy(KFKI, Budapest, Hungary), A. Tanskanen(VTT, Espoo, Finland), F.
Wasastjerna(VTT, Espoo, Finland), R. Kyrki-Rajamaki(VTT, Espoo, Finland), Y. Oka(U.
of Tokyo, Tokyo, Japan), S. Koshizuka (U. of Tokyo, Tokyo, Japan), C. Broeders(FZK,
Karlsruhe, Germany), A. Bergeron(CEA, Saclay, France), E. Kiefhaber(FZK, Karlsruhe,
Germany), D. Sruwe(FZK, Karlsruhe, Germany), P. Rau(Framatome ANP, Erlangen,
Germany), X. Cheng(FZK, Karlsruhe, Germany), T. Schulenberg(FZK, Karlsruhe,
Germany), V. Sanchez(FZK, Karlsruhe, Germany) “Applicable Neutronics/Thermalhydraulics codes, Cross Section Data Base and Neutronics Test Requirements”,
HPLWR-D5, September 2002 (RE)
Sanchez, V. H., Hering, W., Investigations of the appropriateness of RELAP5/MOD3 for
safety evaluations of an innovative reactor operated at thermodynamically supercritical
conditions. Forschungszentrum Karlsruhe Report FZKA-6947, 2002.
D. Squarer (FZK, Karlsruhe), T. Schulenberg (FZK, Karlsruhe), D. Struwe (FZK,
Karlsruhe),Y. Oka (Univ. of Tokyo), D. Bittermann (Framatome ANP, Erlangen), N. Aksan
(PSI, Villigen), C. Maraczy (KFKI, Budapest), R. Kyrki-Rajamaki (VTT, Espoo), A. Souyri
(EDF, Paris), P. Dumaz (CEA, Cadarache), “High Performance Light Water Reactor
(HPLWR)”, Journal of Nuclear Engineering and Design (to be published in 2002)
Squarer, D., “Minutes of The Fifth HPLWR Project Meeting of July 29-31, 2002 at FZK –
Karlsruhe, Germany”, Forschungszentrum Karlsruhe, Germany, HPLWR-M 11,
September 2002 (RE)
Bittermann, D. “General Plant Characteristics”, Framatome ANP, Germany, HPLWR-D
03, August 2002 (RE)
Leistikow, S. (Consultant, Karlsruhe, Germany), Konys, J. (FZK-Karlsruhe, Germany),
Ehrlich, K. (Consultant, MCS-Karlsruhe, Germany), “Available Metallurgical and Process
Parameters/Treatments for Reduction of General Corrosion/Stress Corrosion Cracking
for In- and Ex-vessel Application”, HPLWR-D 9/10, August 2002 (PU)
Aksan, N. (PSI, Switzerland), Bittermann, D., (Framatome ANP, Germany), Squarer, D.
(FZK, Karlsruhe, Germany), “Potential Safety Features of the HPLWR and General
Application of Some Safety Requirements”, HPLWR-D 6, September 2002 (RE)
D. Squarer (FZK, Karlsruhe, Germany), D. Bittermann (Framatome ANP, Erlangen,
Germany), Y. Oka (U. of Tokyo, Tokyo, Japan), P. Dumaz (CEA, Cadarache, France), G.
Rimpault (CEA, Cadarache, France), R. Kyrki-Rajamaki (VTT, Espoo, Finland), K. Ehrlich
(FZK-MCS, Karslruhe, Germany), N. Aksan (PSI, Würelingen, Switzerland), C. Maraczy
(KFKI, Budapest, Hungary), A. Souyri (EdF, Chatou, France), “Summary Report Of The
HPLWR Project”, HPLWR-D 13, October 2002 (RE)
Squarer, D (FZK, Karlsruhe, Germany), “HPLWR Management Report No. 4”, HPLWRP4, October, 2002 (RE)
C.H.M. Broeders, V. Sanchez, E. Stein, A. Travleev, "Validation of Coupled Neutron
Physics and Thermohydraulics Analysis for HPLWR“, FZKA 6742 (2002)
Technological Implementation Plan (TIP) of the HPLWR-Final Version, September, 2002
Leppanen, J, Tanskanen, A., Kyrki-Rajamaki, R., “Feasibility of the TRAB code for
HPLWR reactor dynamics calculations”, VTT Processes, Project Report YR-PR-14/02,
August 2001