HPLWR – D 13 HPLWR Contract No. FIKI-CT-2000-00033 Page 1 of 62

n / EUROPEAN COMMISSION
5th EURATOM FRAMEWORK PROGRAMME 1998-2002
KEY ACTION : NUCLEAR FISSION

HIGH PERFORMANCE LIGHT WATER REACTOR (HPLWR)

CONTRACT N° FIKI-CT-2000-00033

SUMMARY REPORT OF THE HPLWR PROJECT

(HPLWR Deliverable D13)

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)

Dissemination level :
RE: restricted to a group specified by the partners of the HPLWR project

October, 2002 HPLWR – D 13

TABLE OF CONTENTS

Summary

1.Work Package I – Evaluation of current status and definition of basic requirements

1.1Objectives

1.2Description of Work

1.3 Deliverables and milestones

1.4 Methodology

1.5 Review of other concepts

1.6 Design requirements

1.7 Proposed design

1.7.1Primary System

1.7.2Containment and Safety concept

1.7.3Balance of plant (BOP)

Table 1.1- Proposed Characteristics of the HPLWR Power Plant

1.8References

2.Work Package II- Core design and theoretical analyses

2.1 Objectives

2.2 Description of Work

2.3 Deliverables and Milestones

2.4Methodology

2.5 Results of benchmark problem

2.5.1 2-D subassembly calculations

2.5.2Core calculations

2.5.3Conclusions on computer codes for HPLWR

2.6Shortcomings And Proposed Modifications To The Fuel Assembly

2.6.1Design criteria

2.6.2Cladding materials

2.6.3Fuel assembly proposals

2.6.4Solid moderator

2.7Conclusions

2.8References

3. Work Package III – Reactor Safety and Deteriorated Heat Flux

3.1Objectives

3.2Description of Work

3.3 Deliverables and milestones

3.4 Assessment of Required Safety Features

3.4.1Containment Concept and passive safety features

3.4.2Safety Concept

Table 3.1 – Proposed passive and active safety systems for the HPLWR

3.5General Application of Some Safety Requirements

3.5.1European Utility Requirements for Safety versus HPLWR

3.5.2Generation IV Technology Goals in The Safety and Reliability Area versus HPLWR

3.6Deteriorated Heat Transfer in Supercritical Water

3.7Proposed Design

3.8Preliminary Transient Safety Analyses of the HPLWR

3.8.1Use of RELAP5/Mod3 Computer Code

3.8.2Use of CATHARE 2 Computer Code

3.8.3Use of TRAB Computer Code

3.8.4Future work

3.9Conclusions

3.10References

4.Work Package IV-Summary Report on Material Selection and available Treatments for Reduction of Corrosion in HPLWR components.

4.1Objectives

4.2Description of work

4.3Deliverables and Milestones

Table 4-1:HPLWR ”Reference Design” Data for in-core, RPV, and ex-core components

Table 4-2: Estimated maximum temperatures for different materials for the condition of RM/45,000 h at 100 MPa and 200 MPa respectively

4.4References

5.Work Package V – Economics

5.1 Objectives

5.2 Description of Work

5.3 Deliverables and Milestones

5.4Review of other economic studies

5.4.1SCLWR economic study

5.4.2Comparison with other existing concepts

5.4.3GE’s Advanced BWR for improved economics

5.4.4DOE’s Near Term Deployment Economics

Table 5.1 Cost comparison of ALWR and gas turbine plants

5.5Considerations concerning the economics of a HPLWR plant

5.5.1 General

5.5.2 Fuel Cycle Considerations

5.5.3Specific evaluation of HPLWR electricity generation costs

5.4Conclusions

5.5 References

6.Conclusions

Appendix A : List of HPLWR Reports, Minutes and Memos

LIST OF FIGURES

Figure 1. 1 -Example of Hexagonal Fuel Assemblies

Figure 1. 2- Reactor Pressure Vessel (RPV) and Arrangement For Inlet and Outlet Nozzles

Figure 1. 3- HPLWR Containment for a 1000 MWe Plant

Figure 1. 4- Schematic of the HPLWR Circuit Diagram

Figure 3. 1– Containment and Primary Circuit Concept for the HPLWR

Figure 4- 1A comparison of oxidation and spallation for ferritic and austenitic steels at 600°C. Metal loss is half of the oxide thickness.

Figure 4- 2Ultimate tensile strength RM for selected alloys as a function of temperature

Figure 4- 3Creep-rupture strength RM/45,000h for selected alloys

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

Figure 5. 1Cost comparison of Advanced LWR (ALWR) with Combined Cycle Gas Turbine (CCGT) and Gas Turbine (GT) (DOE Near-Term Deployment, October 2001)

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 VI- Project 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 advancedLWRs 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 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 HPLWRwere 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 andpreliminary 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 sub-channels. 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, pressurizer). 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, 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.

1. Work Package I – Evaluation of current status and definition of basic requirements

D. Bittermann(Framatome ANP), D. Squarer(FZK), P. Dumaz(CEA), R. Kyrki-Rajamaki(VTT), C. Maraczy(KFKI), A. Souyri(EdF), N. Aksan(PSI), Y. Oka(U of Tokyo)

1.1Objectives

-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

1.2Description 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

1.3 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 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
  • 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.