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1 A I Effective December 6, 2006, this report has been made publicly available in accordance with Section 734.3(b)(3) and published in accordance with Section of the U.S. Export Administration Regulations. As a result of this publication, this report is subject to only copyright protection and does not require any license agreement from EPRI. This notice supersedes the export control restrictions and any proprietary licensed material notices embedded in the document prior to publication. Program on Technology Innovation: Nuclear Power Generation Technologies Current Status and Trends I C E N S E D L L M A E T R Place Image Here

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3 Program on Technology Innovation: Nuclear Power Generation Technologies Current Status and Trends Final Report, June 2007 EPRI Project Manager G. Ramachandran ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California PO Box 10412, Palo Alto, California USA

4 DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM: (A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT. ORGANIZATION(S) THAT PREPARED THIS DOCUMENT Sargent & Lundy LLC NOTICE: THIS REPORT CONTAINS PROPRIETARY INFORMATION THAT IS THE INTELLECTUAL PROPERTY OF EPRI. ACCORDINGLY, IT IS AVAILABLE ONLY UNDER LICENSE FROM EPRI AND MAY NOT BE REPRODUCED OR DISCLOSED, WHOLLY OR IN PART, BY ANY LICENSEE TO ANY OTHER PERSON OR ORGANIZATION. NOTE For further information about EPRI, call the EPRI Customer Assistance Center at or askepri@epri.com. Electric Power Research Institute, EPRI, and TOGETHER SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc. Copyright 2007 Electric Power Research Institute, Inc. All rights reserved.

5 CITATIONS This report was prepared by Sargent & Lundy LLC 55 East Monroe Street Chicago, IL Principal Investigators B. Andrews D. Demoss B. Gogineni C. Launi This report describes research sponsored by the Electric Power Research Institute (EPRI). The report is a corporate document that should be cited in the literature in the following manner: Program on Technology Innovation: Nuclear Power Generation Technologies: Current Status and Trends. EPRI, Palo Alto, CA: iii

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7 REPORT SUMMARY The United States and other countries are currently planning to expand their nuclear power electrical generation base in order to provide energy security and price stability while reducing greenhouse gas emissions. Since the existing fleet of nuclear plants was built during or before the 1970s, new plants will incorporate more advanced designs. This report documents the current status and potential for advanced nuclear power technology development and/or commercialization over the next 5 to 15 years. Background There are currently 104 nuclear power reactors licensed to operate in the U.S., though only 103 are now operational. Because each of these reactors is fully licensed and meets the U.S. safety standards, a potential builder of a new nuclear power reactor might want to replicate one of these designs for future construction. However, this is not realistic because existing, operating nuclear power reactors in the United States were initiated during or before the 1970s. Technology has progressed and any future construction is likely to incorporate the more advanced designs intended to better meet today's commercial and safety criteria. Objectives To document the current status and potential for advanced nuclear power technology development and/or commercialization over the next 5 to 15 years. Approach To gather technical and cost information, the project team contacted the vendors of six advanced nuclear power technologies: The General Electric Advanced Nuclear BWR (ABWR) The General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) with passive safety systems Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR-1000) PBMR (Pty) Ltd. (South Africa), Pebble Bed Modular Reactor (PBMR) AREVA NP (formerly Framatome ANP), an AREVA and Siemens Company Large Advanced Evolutionary Nuclear Reactor (U.S. EPR not the Finnish or French design) The Westinghouse AP1000, Advanced Nuclear Plant v

8 Each vendor had an opportunity to comments on a draft copy of the section of the report that discussed their technology. The team developed cost estimates for the six technologies from the best available published information the reactor vendors did not provide cost information or endorse the estimates. Results The report provides a general plant description for six advanced nuclear power technologies and estimates costs, in 2006 U.S. dollars, for the construction and operation of each technology. Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. The report summarizes vendor deployment plans and discusses current installations under construction and how building a particular technology in the United States would differ from building a similar unit outside of the United States. A reference section lists internet addresses for many of the references used in compiling the report. These addresses will direct the reader to the location where additional and more detailed information is available. EPRI Perspective The descriptions of the potential technologies and the development of technology supporting data for these nuclear power technologies provided in this report serve as a reference for potential owners and operators of companies interested in considering the option of future nuclear generation. It is not the intent of this report to provide a side-by-side comparison between the currently available nuclear power advanced technologies, but to provide general information on the technologies to inform the reader as to what is available. Likewise, this report does not endorse one technology over another. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline. Keywords Nuclear power Advanced nuclear power technologies Generation III reactors BWRs Pebble bed modular reactor (PBMR) vi

9 ABSTRACT This report provides plant descriptions, estimated cost, and associated technical information that was publicly available as of the Fall/Winter of 2006 on potential advanced nuclear electric power generation technologies. Information on the current status and potential for development and/or commercialization over the next 5-15 years of six nuclear power technologies is summarized. The objective of this report is to develop material that will serve as a reference on the current status of nuclear technology and provide information relevant to planning, engineering, and project development. vii

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11 CONTENTS 1 INTRODUCTION Background Purpose and Goals Report Structure Executive Summary ASSUMPTIONS / LIMITATIONS TECHNOLOGY DESCRIPTION General Electric (GE) Advanced Nuclear BWR (ABWR) Plant Description Design Improvements of the ABWR Estimated Cost Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) With Passive Safety Systems Plant Description Design Improvements of the ESBWR Estimated Cost Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) Plant Description Design Improvements of the ACR Estimated Cost ix

12 3.3.4 Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences PBMR (Pty) Ltd. (South Africa), Pebble Bed Modular Reactor (PBMR) Plant Description Design Enhancement for the PBMR Estimated Cost Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences AREVA NP, an AREVA and Siemens Company Large Advanced Evolutionary Nuclear Reactor (U.S. EPR Not the Finnish or French Design) Plant Description Design Improvements of the U.S. EPR Estimated Cost Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences AP1000 Advanced Nuclear Plant Plant Description Design Improvements of the AP Estimated Cost Deployment Plans Summary Installations Under Construction or Planned U.S. Design Differences CONCLUSIONS REFERENCES x

13 LIST OF FIGURES Figure 3-1 GE ABWR Simplified Flow Diagram [5] Figure 3-2 ESBWR Safety System Configuration (not to scale) [25] Figure 3-3 Typical ESBWR Power Block Arrangement 26] Figure 3-4 ESBWR Reactor Pressure Vessel System Key Features [24] Figure 3-5 ESBWR Reactor Pressure Vessel Natural Circulation Process [24] Figure 3-6 Overall ACR-1000 Plant Simplified Flow Diagram (Provided by AECL) [32] Figure 3-7 CANFLEX -ACR Fuel Bundle (Provided by AECL) [32] Figure 3-8 ACR-1000 Nuclear Systems Schematic (Provided by AECL) [32] Figure 3-9 ACR-1000 Reserve Water System (Provided by AECL) [32] Figure 3-10 Simplified Schematic Diagram of the PBMR Main Power System (Provided by PBMR (Pty) Ltd) [35] Figure 3-11 Physical Layout of the PBMR Main Power System (Provided by PBMR (Pty) Ltd) [35] Figure 3-12 Schematic Diagram of the PBMR Fuel Handling System During Normal Operation (Provided by PBMR (Pty) Ltd) [35] Figure 3-13 Physical Layout of the PBMR Fuel Handling and Storage System (Provided by PBMR (Pty) Ltd) [35] Figure 3-14 PBMR Fuel Element System (Provided by PBMR (Pty) Ltd) [35] Figure 3-15 Sample Plant Layout of the U.S. EPR (Provided by AREVA NP) Figure 3-16 Simplified Process Flow Diagram for the EPR (Provided by AREVA NP) [50] Figure 3-17 Aeroball System Schematic for the U.S. EPR (Provided by AREVA NP) [55] Figure 3-18 Example of the Aeroball System Configuration in the Reactor [52] Figure 3-19 AP1000 Reactor Coolant System (Provided by Westinghouse Electric Co., LLC.) [63] Figure 3-20 AP1000 Passive Core Cooling System (Provided by Westinghouse Electric Co., LLC.) [61] Figure 3-21 AP1000 Passive Containment Cooling System (Provided by Westinghouse Electric Co., LLC.) [61] xi

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15 LIST OF TABLES Table 1-1 Confidence Rating Based on Technology Development Status Letter Rating Key Word Description Table 1-2 Confidence Rating Based on Cost and Design Estimate Table 3-1 Technology Basis Data (ABWR) Table 3-2 Technology Monitoring Guide (ABWR) Table 3-3 Technology Development and Assessment (ABWR) Table 3-4 Technology Basis Data (ESBWR) Table 3-5 Technology Monitoring Guide (ESBWR) Table 3-6 Technology Development and Assessment (ESBWR) Table 3-7 Technology Basis Data (ACR-1000) Table 3-8 Technology Monitoring Guide (ACR) Table 3-9 Technology Development and Assessment (ACR) Table 3-10 Technology Basis Data (PBMR) Table 3-11 Technology Monitoring Guide (PBMR) Table 3-12 Technology Development and Assessment (PBMR) Table 3-13 Technology Basis Data (U.S. EPR) Table 3-14 Technology Monitoring Guide (U.S. EPR) Table 3-15 Technology Development and Assessment (U.S. EPR) Table 3-16 Technology Basis Data (AP1000) Table 3-17 Technology Monitoring Guide (AP1000) Table 3-18 Technology Development and Assessment (AP1000) xiii

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17 1 INTRODUCTION 1.1 Background Many countries including the United States (U.S.) are currently making plans to expand their nuclear power electrical generation base. The use of additional nuclear power electrical generation would help these countries meet their growing energy needs and would help to provide energy security and price stability while reducing greenhouse gas emissions. There are currently 104 nuclear power reactors licensed to operate in the U.S., though only 103 are now operational. Because each of these reactors is fully licensed and meets the U.S. safety standards, a potential builder of a new nuclear power reactor might want to replicate one of these designs for future construction. However, this is not realistic because existing, operating nuclear power reactors in the U.S. were initiated during or before the 1970s. Technology has progressed and any future construction is likely to incorporate the more advanced designs intended to better meet today's commercial and safety criteria. Generation II reactors are typified by the present U.S. reactor fleet and most reactors in operation elsewhere. Generation III (and III+) are the Advanced Reactors. Generation III reactors include the advanced boiling water reactor (ABWR), the pressurized water reactor System 80+, and the AP600 passive-design reactor. These designs were developed in the U.S. in cooperation with overseas organizations and certified by the U.S. Nuclear Regulatory Commission (NRC) in the 1990s. Generation III+ reactors are reactors that can be deployed (construction started) by They have been under development during the 1990s and are in various stages of design and implementation now. They include the ESBWR, ACR-700, ACR-1000 (ACR activities outside the U.S), U.S. EPR, and the AP1000, which has been certified by the NRC. Generation IV nuclear energy systems are future, next-generation technologies that will compete in all markets with the most cost-effective technologies expected to be available over the next three decades. Comparative advantages for Generation IV reactors include reduced capital cost, enhanced nuclear safety, minimal generation of nuclear waste, and further reduction of the risk of weapons materials proliferation. Generation IV systems are intended to be responsive to the needs of a broad range of nations and users. The purpose of Generation IV systems is to develop nuclear energy systems that would be available for worldwide deployment by 2030 or earlier. The pebble-bed modular reactor (PBMR) would be considered to be a Generation IV reactor. This report provides summary information on six (6) advanced nuclear power technologies. [66, 67, 68] 1-1

18 Introduction 1.2 Purpose and Goals This report documents the current status and potential projections for advanced nuclear power technology development and/or commercialization over the next 5 to 15 years. The descriptions of the potential technologies and the development of technology supporting data for these nuclear power technologies provided in this report serve as a reference for potential owners and operators of companies interested in considering the option of future nuclear generation. It is not the intent of this report to provide a side-by-side comparison between the currently available nuclear power advanced technologies, but to provide general information on the technologies to inform the reader as to what is available. Likewise, this report does not endorse one technology over another. This report addresses issues and activities related to gaining an understanding of new nuclear generation technologies as they relate to planning, engineering, project development and preparation for commercial deployment. Literature searches from publicly available information and databases such as INPO, EPRI, NEI, NRC, the internet, and vendor literature were performed to compile the data presented in this report. This report addresses the following advanced nuclear power technologies: 1. The General Electric Advanced Nuclear BWR (ABWR) 2. The General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) with passive safety systems 3. Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR-1000) 4. PBMR (Pty) Ltd. (South Africa), Pebble Bed Modular Reactor (PBMR) 5. AREVA NP (formerly Framatome ANP), an AREVA and Siemens Company Large Advanced Evolutionary Nuclear Reactor (U.S. EPR not the Finnish or French design) 6. The Westinghouse AP1000, Advanced Nuclear Plant 1.3 Report Structure For each of the aforementioned nuclear power technologies, a general plant description is provided with a simplified process flow diagram. This information is included to provide an understanding of the nuclear generation technologies and the engineering, and project development that would be needed to support the construction of a particular nuclear power technology. There is a section on design improvements related to the nuclear power technology and a section for estimated costs for the construction and operation of the described nuclear technology. The dollars values stated in this report should be assumed to be 2006 U.S. dollars unless otherwise noted. This is followed by sections that summarize vendor deployment plans, governmental agency technology evaluations of the nuclear technology or the type of reports that are being developed and by which governmental agency (U.S. NRC, Canadian NSC, etc.). There is a discussion on current installations under construction or planned for the nuclear power technology in question and a discussion on how building a particular nuclear technology in the U.S. would differ from building a similar unit outside the U.S. At the end of the report is the reference section that lists internet addresses for many of the references used in compiling the 1-2

19 Introduction data summarized in this report. If the reader requires more information on a particular nuclear technology, these internet addresses will direct the reader to the location where additional and more detailed information is available on each of the nuclear technologies discussed in this report. Each of the aforementioned nuclear power technologies vendors were contacted and requested to supply both technical and cost information for there respective nuclear power technology. Also each of the aforementioned nuclear power technologies vendors were given an opportunity to provide comments on a draft copy of the section of this report where there respective nuclear power technology is discuss. The degree of technology information transfer varied among the vendors based on current design progress. Little or no cost information was provided from the nuclear power technologies vendors because of the early status of the U.S. plant designs and no construction releases in the U.S. Cost and Technical Data Confidence Rating Two rating systems are used in this report to give an overall confidence level to the data presented. One system is based on a technology s development status; the other is based on the level of effort expended in the design and cost estimate. The rating system shown in Table 1-1 below indicates the developmental status for the technology. Table 1-1 Confidence Rating Based on Technology Development Status Letter Rating Key Word Description Letter Rating Key Word Description A B Mature Commercial Significant commercial experience (several to many operating commercial units) Recent commercial experience or some commercial experience C Demonstration Concept verified by integrated demonstration unit D Pilot Concept verified by small pilot facility E Laboratory Laboratory Concept verified by laboratory studies and initial hardware development F Idea No system hardware development Some degree of uncertainty is generally expected in cost and performance data. Because new technologies do not have a history of construction or operating costs, only estimates can be used. The rating system shown in Table 1-2 below indicates the level of effort involved in the design and cost estimate. 1-3

20 Introduction Table 1-2 Confidence Rating Based on Cost and Design Estimate Letter Rating Key Word Description A Actual Data on detailed process and mechanical designs or historical data from existing units B Detailed Detailed process design C Preliminary Preliminary process design D Simplified Simplified process design E Goal Technical design/cost goal for value developed from literature data To ensure a common basis for comparison, the capital cost of nuclear electric generation technologies are expressed in dollars per kilowatt of capacity. The capital costs used in such comparisons are called overnight capital costs. Overnight capital costs assume that the nuclear plant is built overnight and thus do not include interest and financing costs. There are many risks (uncertainties) in calculating the allowance for funds used during construction (AFUDC). These risks include the average debt rates, cash-flow models, interest rates, and the length of construction. Owner s costs include the COL (Combined Operating License) application preparation and submittal, land procurement, transmission upgrades, security infrastructure, emergency preparedness, meteorological monitoring, testing systems within the plant, training a staff (which may take several years while the plant is being built), simulator procurement (costs to build and maintain a simulator of the given reference reactor type could be shared between the owners of that reactor type), site improvement and access roads, various inspections, etc. Risks associated with a new nuclear plant project can be considered in terms of how they affect time-related costs that are impacted by delays in the project schedule, such as interest payment on funds used during construction, and non-time-related costs, such as higher-than-expected material or labor costs. The primary risks that affect the costs associated with the construction of a new nuclear plant are: Project management Changes in the certified design Changes in digital controls Availability of skilled engineering and construction personnel/labor Capacity factor Licensing processes: 10 CFR Part 52 & 50 Availability of key equipment 1-4

21 Introduction Effectiveness of the modularization construction process Effectiveness of construction planning/assistance software: multi-d CAD-CM, advanced digital info systems Escalation in material costs Availability of financial incentives Safety goal standardization Design standardization within families of plants Radioactive waste disposal Lack of effective project management represents the greatest risk to overall nuclear project costs in terms of both likelihood and severity. Although the reference plants provide a level of confidence in the technical design and construction approach to the facility, the applicability of project management experience is not clear. The available resources of nuclear suppliers are expected to be in great demand in the future. Utility management will need to ensure that the proposed project team for the construction of a new nuclear power plant has acceptable expertise, commitment and support from the parent companies. Changes in certified design, digital controls and the availability of skilled labor for nuclear plant construction have medium to high severity with medium likelihood. The process of submittal and approval of changes to Certified Design is not tested. Examples are upgrades like digital controls and site specific limitations that may not match certified design. In addition to the risks related to the licensing process, changes in the digital controls have technical risks. Control room work is often a critical path for a project during the latter stages of construction. Implementing a new, software based and licensing dependent control system will result in the potential for commissioning delays after construction is substantially complete. Skilled labor in various engineering disciplines and crafts (welders, electrical, etc) will likely be limited due to other nuclear projects and non-nuclear projects that compete for resources; though, as the demand becomes more certain, supply will expand to meet the increase in demand. An example of how supply meets demand is that BWX Technologies Mount Vernon, Indiana facility, which once produced large, heavy components for the nuclear industry and recently had reinstated its American Society of Mechanical Engineers N-stamp certification a crucial requirement for fabricating commercial nuclear-grade components. The BWXT facility plans to revive manufacturing of large, heavy nuclear components if new plant orders materialize. It is currently the only such facility in the U.S. with such capabilities. Meantime, it will manufacture replacement components, including reactor vessel heads, for existing plants. It has an order to build reactor vessel heads for Pacific Gas & Electric s Diablo Canyon. [69] Several risks associated with nuclear plant projects have a low likelihood of occurring, but medium to high impact (increased cost). The highest potential impact of these low probable risks is the performance of the unit once completed, as measured by capacity factor. The proposed plants should be able to achieve an industry capacity factor goal of above 90%. However, in the past plants have encountered unexpected equipment or regulatory problems and have been forced into extended (6 months to 3 years) outages. This potential is very real with a new plant design. 1-5

22 Introduction Although the new 10 CFR Part 52 licensing process is intended to minimize the risk of project delays due to licensing issues, 10 CFR Part 52 has not been used except for Early Site Permit applications where delays have been encountered. If equipment is delivered late, it has the potential to become a critical path item and delay the project. This becomes more critical for the equipment procured later because a larger portion of the cost will be incurred at the time of the delay. The quantification of benefits for modularization of nuclear plants constructed in the U.S. is not yet clearly established. Other risks associated with a nuclear project have medium likelihood, but low impact. Radioactive waste disposal is an example of such a risk. Although there are considered to be few environmental risks associated with the construction and operation of a nuclear power plant, due to low emissions and overall performance of the current U.S. domestic fleet, radioactive waste disposal does represent an environmental risk for new plant designs. The low-level and highlevel waste disposal options and storage requirements for the new units do not differ from operating U.S. plants in a significant way. The plants are planned for a limited amount of onsite storage, after which the material would be sent to a long-term disposal facilities. The availability of disposal options for both low-level and high-level radioactive waste, and associated long term storage requirements, are uncertain. This could require additional investment in short term storage. The risk is considered to have a medium likelihood of occurring, but has a low severity relative to overall project costs. Nevertheless, it is imperative that the Yucca Mountain repository for spent fuel be licensed and placed into operation. Some states require that action before new nuclear plants can be built in their state. The utility industry (NEI) has stated that longer term sustained expansion of nuclear power depends on completing that achievement. The primary impact of the risks related to the construction of a nuclear power plant is increased capital costs due to construction delays and uncertainties in initial cost estimates. Some of the risks can be mitigated through effective planning and contracting strategies. However, not all of the risk is expected to be wrapped into the project engineering, procurement and construction contracts as all the risk is not within the contractors control, such as licensing and availability of material and qualified personnel. A significant number of the causes of the delays and cost overruns associated with building the current U.S. fleet of nuclear plants have been addressed by structural changes in the design and licensing process, such as the development and implementation of the 10 CFR Part 52 COL and design certification processes. These risks are not project specific, and so mitigation strategies need to be aligned with industry programs and efforts to build new nuclear plants. Some of the risks, however, are project specific, although they are impacted by industry issues, and can be expected to be borne or shared by the contractors. For example, risks related to effective project management, modularization construction techniques, and effective implementation of approved design changes are more within the contractors control than generic licensing issues. Accordingly, corresponding mitigation strategies should be project specific and can be addressed during the project development, planning and contracting stages of the project. 1-6

23 Introduction 1.4 Executive Summary This report does not endorse one nuclear power technology over another. The following table provides a summary of estimated costs and associated technical information for the six nuclear power technologies that are discussed in this report. These cost estimates were developed from the best available published information and were not directly provided by the reactor vendors and are not endorsed by the reactor vendors. The cost estimates were developed based on nuclear plant cost information that was publicly available and industry cost factors and studies. No guarantee is provided as to the accuracy of these values. For example, these values would change as a result of increasing material, equipment and labor prices. Summary of Selected Technology Basis Data* Plant Type Data Type General Electric ABWR General Electric ESBWR AECL ACR-1000 PBMR (South Africa) AREVA NP, U.S. EPR Westinghouse AP1000 Fuel Type Enriched Uranium Dioxide GE14 (10x10 fuel assembly) Enriched Uranium Dioxide 4.2% GE14 (10x10 fuel assembly) Slightly Enriched Uranium Dioxide - Average <2.0 wt% 235 U. 43- element CANFLEX - ACR fuel bundle design Enriched Uranium Dioxide, Ceramic spheres covered in graphite Enriched Uranium/MOX 17 x 17 lattice Enriched Uranium Dioxide Full MOX core capability 17x17 fuel assemblies. Fuel Cycle 24-month refueling interval or An initial 12- month cycle followed by two 18-month cycles with subsequent cycles of 24- months refueling intervals. 24 month cycle 25 day outage duration On line refueling Maintenance outage of 21 days every 3 years and a mid-life outage for pressure tube replacement. On line refueling 12-month to 24-month cycles 18 months 17 day refueling outage Plant Size**, MWe Plant Size, Net MWe 1,438 (gross) 1,600 (gross) 1,165 (gross) 175 (gross) 1,600 (gross) 1,200 (gross) 1,371 1,535 1, ,600 1,

24 Introduction Summary of Selected Technology Basis Data* (continued) Plant Type Data Type General Electric ABWR General Electric ESBWR AECL ACR-1000 PBMR (South Africa) AREVA NP, U.S. EPR Westinghouse AP1000 Available for Orders, Year Now Now Now 2012 Now Now First Commercial Service (Technology Year) 1996 Japan (Finland design) 2015 (U.S. EPR) 2015 U.S. 2013/2014 China Total Plant Cost $2,683 million ($1957/kWe) $2,420 million ($1577/kWe) $2,610 million ($2,406/kWe) $195 million ($1219/kWe) $3,225 million ($2,016/kWe) $2,020 million ($1812/kWe) Total Plant Investment (includes AFUDC) $3,106 million ($2265/kWe) $2,800 million ($1824/kWe) $2,970 million ($2,737/kWe) $210 million ($1,313/kWe) $3,750 million ($2344/kWe) $2,270 million ($/kwe) Total Owner Costs*** $310 million ($226/kWe) $280 million ($182/kWe) $300 million ($276/kWe) $20 million ($125/kWe) $375 million ($234/kWe) $230 million ($2036/kWe) Fixed $/MW-yr (Staff of 701 at a Greenfield site) (staff of 701 at a Greenfield site) $9.81 (ACR-700 greenfield twin unit with 761 staff) Information not available Information not available 11.40/Net MWe (698 staff at a Greenfield site) 2003 dollars Preconstruction, License & Design Time, Years 4 to 5 4 to to 6 years (for the U.S.) 4 4 to 5 Idealized Plant Construction Time, Years 3.3 (40 months) 3.25 (39 months) 3.5 (42 months) 2 years for the first module plus 6 months for follow-on modules. 4 (48 months) 3 (36 months) Unit Life, Years

25 Introduction Summary of Selected Technology Basis Data* (continued) Plant Type Data Type General Electric ABWR General Electric ESBWR AECL ACR-1000 PBMR (South Africa) AREVA NP, U.S. EPR Westinghouse AP1000 Estimated Land Area Needed (acres) Excluding cooling ponds 270 (minimum) The EAB would be about 0.5 miles from the containment. 270 (minimum) The EAB would be about 0.5 miles from the containment. 260 (minimum) The EAB would be about 0.3 miles from the containment. 125 (minimum) The EAB would be about 0.25 miles from the reactor building. 300 (minimum) The EAB would be about 0.5 miles from the containment. 370 (minimum) The EAB would be about 0.5 miles from the containment. Estimated Water Requirements (gpm) 41,000 41,000 28,000 Once through cooling: (~38,000 gpm for 400 MWt) 41,000 (estimated) 32,000 * For more information see the section that discusses the nuclear power technology of interest. ** Electrical output values can vary depending on the Turbine Island design and site-specific conditions (To be confirmed in COL phase). *** Based on 10% of the total plant cost The fuel channels are designed to achieve a 30-year operating life. Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline. 1-9

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27 2 ASSUMPTIONS / LIMITATIONS 2.1 This assessment of nuclear power generation technologies is not a cost estimating study. 2.2 All values listed in this report are approximate values. Values like electrical output, power output (gross or net) are dependent on site conditions. 2.3 The information contained in this report reflects the details of the nuclear technologies as they were known at the time this report was generated (Fall/Winter 2006). 2.4 The level of detail presented in this report for each nuclear technology is based on the amount of information that was readily available when the literature searches were conducted and the vendor information provided. 2.5 The reference section lists internet addresses for many of the references used in compiling the data summarized in this report. Internet sites are continually updated and changed; therefore, the information taken from a particular referenced internet site may be changed at any time and therefore, the new or different information contained in the internet site would not be reflected in this report. 2.6 All five reactor vendors were reluctant to supply detailed cost estimates for their associated plant designs. Pricing information was obtained from published literature data. 2.7 Westinghouse Electric Company, LLC did not provide comments on nor concurrence with the information provided in this report, either technically or commercially. 2.8 AREVA NP, did not provide any written comments on the information provided in this report, either technically or commercially. AREVA NP considers the pricing information for the U.S. EPR is currently proprietary information. As such, the values in this report represent independent efforts and do not reflect any position on the part of AREVA NP. 2.9 Atomic Energy of Canada Limited (AECL) provided technical comments on the information provided in this report for the advanced CANDU Rector (ACR-1000) design PBMR (Pty) Ltd. (South Africa) provided technical comments on the information provided in this report for the Pebble Bed Modular Reactor (PBMR) design Additional information is available in the URDs (U.S. Utility Requirements Documents) and the URD is not part of this assessment of nuclear power generation technologies. 2-1

28 Assumptions / Limitations 2.12 A lot of work is going on in areas of chemistry controls to remedy some of the earlier corrosion-related reliability problems with the BWR and the intent of this report is for planning purpose and will not address these issues in detail. Many of these improvements have been made on the current operating plants This report does not include the levelized cost of electricity for the nuclear power generation technologies discussed All dollars reported should be assumed to be 2006 dollars unless stated differently. All costs are based on 2004 dolor, escalated to 2006 dollars inflation rat of 5.6%. The high inflation rates of maternal equipment and labor from have not been included in the estimate Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline. 2-2

29 3 TECHNOLOGY DESCRIPTION 3.1 General Electric (GE) Advanced Nuclear BWR (ABWR) General Electric s (GE) ABWR (Advanced Boiling Water Reactor) nuclear plant has been designed and licensed in Japan and Taiwan and granted standard design certification in the U.S. (10 CFR 52, Appendix A). In general, BWRs use low-enriched uranium as fuel to produce heat and boil water in the reactor s core. The resulting steam is used to directly drive a turbine, which is coupled to an electrical generator to produce electrical energy Plant Description The ABWR is a single cycle, forced circulation BWR. In a typical boiling water reactor, the reactor core creates heat and the single coolant cycle loop delivers steam to the turbine and returns water to the reactor core to cool it. The cooling water is force-circulated by electrically powered motor driven pumps. GE developed the 1,438 MWe (nominal) ABWR in cooperation with the Tokyo Electric Power Company, Hitachi, and Toshiba. The reactor is light water-cooled and moderated and utilizes enriched uranium as fuel. The ABWR incorporates design features proven in many years of worldwide BWR operating experience, along with advanced features such as vessel-mounted reactor recirculation pumps, fine motion control rod drives and a stateof-the-art digital, multiplexed, fiber-optic control and instrumentation system. Figure 3-1 provides a simplified flow diagram of the GE ABWR. The reactor and turbine building are arranged "in-line." At multiple unit sites, none of the major facilities are shared with the other units. The containment is a reinforced concrete containment vessel (RCCV) with a leak tight steel lining. The containment is surrounded by the reactor building, which doubles as a secondary containment. A negative pressure is maintained in the reactor building to direct any radioactive release from the containment to a gas treatment system. The reactor building and the containment are integrated to improve the seismic response of the building without additional increase in the size and load bearing capability of the walls. [5, 6, 8] 3-1

30 Figure 3-1 GE ABWR Simplified Flow Diagram [5] Design Improvements of the ABWR The ABWR has several advantages over its parent BWR design. These include vessel-mounted recirculation pumps, fine motion control rod drives, improvement to containment design, and an advanced digital and a multiplexed instrumentation and control system. The Toshiba and GE versions of the ABWR are essentially identical, and for the purposes of this report, unless stated, a common version of the ABWR is referenced. 6] Details of the design improvements of the ABWR are: 1. Current BWR designs have hydraulic control rods which are used to shutdown the nuclear reaction in the reactor. In the ABWR design, the control rods are electro-hydraulic (Fine Motion Control Rod Drive). Having an additional drive mechanism reduces the probability of failure of the control rods and improves the plant s ability to follow the electrical load. [7] 2. Compared to the more recent BWRs, safety improvements have been incorporated into the ABWR, also resulting in a more compact design than the current BWRs (i.e. 70% smaller building volume). This decrease in size not only lowers construction time and cost but it further improves plant safety by making the plant more rugged and more immune to earthquakes. [7] 3. One significant safety improvement to the ABWR is that it has three independent and redundant divisions of safety systems that have no cross connections (mechanical or electrical) as in earlier BWR designs. Each division is also physically separated by fire walls and has its own emergency diesel generator, allowing complete isolation for full independent capability of each system. [8] 3-2

31 4. The external recirculation loops have been eliminated from the reactor. The external recirculation pumps and piping have been replaced by ten internal recirculation pumps mounted to the bottom head. These Reactor Internal Pumps (RIPs) are improved versions of those used in Europe for which there is over 1000 pump-years of operating experience. The reliability and durability of these pumps has proven to be so good that it is estimated that only two will be removed for servicing during an outage. The RIP motors are continuously purged with clean water to keep crud from settling into them from the vessel, so that radiation levels around the pumps are vastly reduced. [8] 5. The reactor pressure vessel (RPV) of the ABWR (U.S. standard design) is 21 meters (~69 ft.) high and 7.1 meters (~23 ft.) in diameter. Much of the vessel is made from a single forging, and since the external recirculation loops have been eliminated the vessel has no nozzles below the top of the core greater than 2 inches in diameter. This allows for 50% of the welds and all the piping and associated supports in the primary system to be eliminated. It is important to note that these components served as the largest source of occupational exposure in the current BWR design. [8] 6. The reactor and turbine building are arranged "in-line." In multiple unit plants none of the major facilities are shared with the other units. The containment is a reinforced concrete containment vessel (RCCV) with a leak tight steel lining. The containment is surrounded by the reactor building, which serves as the secondary containment. A negative pressure is maintained in the reactor building to direct any radioactive release from the containment to a gas treatment system. The reactor building and the containment are integrated to improve the seismic response of the building without additional increase in the size and load bearing capability of the walls. [8] 7. All major equipment and components have been engineered with service and maintenance in mind, which will minimize downtime and reduce worker exposure to radiation. [7] Estimated Cost The first two ABWRs, Kashiwazaki Kariwa-6 & 7 were built in Japan and have been operating since 1996 and are designed to have a 60 year life. These GE-Hitachi-Toshiba units cost approximately $2,000/kWe to build, and produce power at approximately 7 cents/kwh. Future ABWR units are expected to cost approximately $1,700/kWe (in 2004 the capital cost was estimated to be $1,600 per kwe, 2003 dollars). The estimated overnight engineering, procurement and construction costs for TVA s Bellefonte site were $1,611/kWe (2004 U.S. dollars) for twin units with a one-year lag for commercial operation. Assuming a $1,720/kWe overnight cost for a 1,438 MWe unit would mean the unit would have an overnight construction cost of $2,473 million. This value could grow to approximately $2,380/kWe once the owner s costs and the AFUDC are included (~$3,420 million), see Table 3-1 for cost estimates (Note: These cost estimates were not directly provided by the reactor vendor and are not endorsed by the reactor vendor. The cost estimates were developed based on the information that was publicly available. No guarantee is provided as to the accuracy of these values. For example, these values would change as a result of increasing material, equipment and labor prices). The December 7, 2006 issue of Platts Nucleonics Week stated that earlier this year (2006), executives at Westinghouse, AREVA NP and General Electric estimated the cost for a new reactor at between $1,600 and $2,000 per installed kilowatt. [9, 10, 11, 12, 13] 3-3

32 Table 3-1 Technology Basis Data (ABWR) Plant Type General Electric Advanced Nuclear BWR (ABWR) Fuel Type Enriched Uranium Dioxide GE14 (10x10 fuel assembly) (Reference Core design) [6, 12] 24-month refueling interval or Fuel Cycle An initial 12-month cycle followed by two 18- month cycles with subsequent cycles of 24- months refueling intervals [12] Number of Units 1 [6] 25 day outage duration [6] 3,992 MWt [12] Unit Size, MW 1,438 MWe (gross) [12] Electrical output values can vary depending on the Turbine Island design and site-specific conditions. Plant Size, Net MWe 1,371 [6, 12] Plant Auxiliaries, MWe 67 [12] (1,438 MWe 1,371 MWe) Plant Capacity Net 95% Available for Commercial Orders, Year Now First Commercial Service (Technology Year) 1996 Japan [6] Hypothetical In-Service Year Design & Cost Estimate Rating Technology Development Rating Plant Location Plant Capital Cost Structures & Improvements month construction time estimated [12] B Commercial B Detailed U.S. 1 Unit $288 million ($210/kWe) [12 adjusted for 2006 dollars] 3-4

33 Table 3-1 Technology Basis Data (ABWR) (continued) Plant Type Reactor Plant Equipment Turbine Plant Equipment Electrical Plant Equipment General Electric Advanced Nuclear BWR (ABWR) $509 million ($371/kWe) [12 adjusted for 2006 dollars] $340 million ($248/kWe) [12 adjusted for 2006 dollars] $142 million ($104/kWe) [12 adjusted for 2006 dollars] $50 million ($36/kWe) for miscellaneous plant equipment Other BOP Facilities Installation General Facilities & Engineering Fee $21 million ($15/kWe) for the Main Condenser Heat Rejection System $183 million ($133/kWe) for construction services, engineering / home office services and field supervision / field office services [12 adjusted for 2006 dollars] $525 million ($383/kWe) manual labor $415 million ($303/kWe) Non manual labor [12 adjusted for 2006 dollars] See Other BOP Facilities above $145 million ($106/kWe) Project & Process Contingency Total Plant Cost Total Cash Expended (Mixed year $) AFUDC (interest during construction) Total Plant Investment (includes AFUDC) Total Owner Costs Total Capital Requirement, Hypothetical In-Service Year (includes AFUDC) 1,144 MWe Plant in 2001 Dollars [10] Use $210 million ($153/kWe) (adjusted for a 1,438 MWe unit and 2006 dollars) [14] $2,683 million with contingency ($2683/kWe) To be completed by EPRI $423 million ($308/kWe) (Based on 8% interest, $67 million per month for 40 months) $3,106 million ($2265/kWe) $310 million ($226/kWe) (Based on 10% of the total plant cost) To be completed by EPRI 3-5

34 Table 3-1 Technology Basis Data (ABWR) (continued) Plant Type General Electric Advanced Nuclear BWR (ABWR) Cost for Hypothetical In-Service Year Operating & Maintenance Costs Fixed $/MW-yr Incremental, mils/kwh: Variable (includes consumables), mills/kwh Net Plant Heat Rate, Btu/kWh (if applicable) Full Load Min Load Average Annual [6] (Staff of 701 at a Greenfield site) 4.64 fuel cost 2.45 variable 1 Nuclear Waste Fee 8.09 total [10] Information not available Information not available Information not available Unit Availability Equivalent Planned Outage Rate, % Information not available Equivalent Unplanned Outage Rate, % 4 (forced outage rate) [6] Equivalent Availability, % Capability Ratio 93% [6] Duty Cycle Minimum Load, % Information not available Preconstruction, License & Design Time, Years 4 to Idealized Plant Construction Time, Years 40 months [12] (43 Months) [6] Unit Life, Years (minimum) Estimated Land Area Needed (acres) Excluding cooling ponds The EAB would be about 0.5 miles from the containment. 3-6

35 Table 3-1 Technology Basis Data (ABWR) (continued) Plant Type General Electric Advanced Nuclear BWR (ABWR) Estimated Water Requirements (gpm) 41, Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. 2. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline Deployment Plans Summary The world s first ABWR has been in commercial operation since November 7, The Kashiwazaki-Kariwa Unit No.6, 1,356 MWe, is owned and operated by the Tokyo Electric Power Company (TEPCO). Kashiwazaki-Kariwa Unit No.7, same capacity as Unit No.6, has been in commercial operation since July 2, 1997 and Hamaoka Unit No.5, 1,325 MWe, has been in commercial operation since January 18, 2005, both are also ABWR designs. Two more ABWR s are nearing completion in Taiwan. 3-7

36 Table 3-2 Technology Monitoring Guide (ABWR) Leading Developers of the Science or Technology Unresolved Issues Changes to Watch For Major Trends Leading Vendors Industrial Firms Nonprofit Organizations Government Organizations R&D Intensity Technologies ABWR Extensive testing and development was conducted to ensure the underlying technology was sound and that the plant would perform as expected TVA 1 DOE 2 EPRI including the BWR Owners Group Bechtel Power Corporation, Global Nuclear Fuel America GE 3, Toshiba, Hitachi [2] Since, 1980 the construction time for a BWR in Japan has decreased steadily. Also, the construction manpower at the job site has been steadily decreasing in Japan. These reductions are due to utilizing design construction technology improvements. [12] Cost saving if safety-related but low safetysignificant components are procured commercially. Reactor designs with fewer moving parts would see less benefit. Would require changes to 10 CFR Part 52 for the new plant licensing process. Enhancement of modularization during construction to reduce the field installation work. [12, 18] There are no major design or engineering gaps to overcome. [10, page 4-1] 1. Tennessee Valley Authority 2. Department of Energy 3. General Electric 3-8

37 Table 3-3 Technology Development and Assessment (ABWR) Features and Characteristics of Technology Major Technical Issues Current Status Forced Circulation BWR Reactor and turbine building are arranged in-line At multiple unit sites, none of the major facilities are shared between the units The ABWR design eliminates the main steam isolation valve leakage control system (MSIVLCS) for the ABWR plant design. Future Considerations or Trends Information not available NUREG-0800 (Section 6.7) and Regulatory Guide 1.96 may be revised to show that the NRC has taken the position that the elimination of the MSIVLCS by taking credit for fission product plate-out and holdup in the main steam lines and the condenser would be acceptable. [19, Appendix I] Key Vendor Activities COL activities in the U.S. Information not available Resource Requirements That Impact Technology Key Business and Market Indicators None COL applications to the NRC that use the ABWR as the reference plant Information not available Start of construction of an ABWR in the U.S. Key technology needs Information not available Information not available Technology Outlook Development Timeframe ABWRs have been build in Japan and are under construction in Taiwan Information not available Research N/A Information not available Development N/A Information not available Demonstration N/A Information not available Projected Commercialization Date 1996 Information not available 3-9

38 The technology evaluation was carried out by the utilities, aided by EPRI and its contractors. The ABWR design is certified by the U.S. Nuclear Regulatory Commission. During the design and construction of Kashiwazaki-Kariwa Units 6 and 7 in Japan, GE pursued Design Certification of the ABWR design in the United States. GE submitted the GE ABWR design to the NRC for design certification under 10 CFR Part 52. The NRC issued a Final Design Approval for the GE ABWR in 1996 and Design Certification in EPRI and its Utility Steering Committee adopted the ABWR design as part of the Advanced Light Water Reactor Program in the mid-80s, and GE completed First of a Kind Engineering (FOAKE) design of the ABWR with EPRI utility member funding cost-shared by DOE in Toshiba participated in GE s activities in the Design Certification Process and FOAKE process and some of the documents were prepared by Toshiba. After the successful completion of construction of the first ABWR in Japan and the issuance of the Design Certification by the NRC, GE succeeded in winning the award for the contract to construct two units of the ABWR at the Lungmen site in Taiwan. The Taiwan Power Company required that the design to be built at the Lungmen site already be licensed in the country of origin and so GE utilized its U.S. NRC Certified ABWR design. GE also applied the design detailing performed under the EPRI FOAKE Program. As a result, the ABWR design being constructed at the Lungmen site is based on U.S. NRC licensing requirements and will be consistent with the U.S. Utility Requirements Document (URD). As a subcontract to GE, Toshiba provided key ABWR equipment including a reactor pressure vessel, reactor internals, and reactor internal pumps (RIP) for the Lungmen Plant. Based on the current design certification, Toshiba and GE developed the plant concept, incorporating lessons learned and technology advancements developed during the Japanese and Lungmen unit design and construction. In Europe, the ABWR was being adapted to European and Finnish requirements in The ABWR was being reviewed in detail by the European Utility Organization (EUR) Organization against European regulatory requirements, and GE was working with Teollisuuden Voima Oy (TVO) to assess the feasibility of the ABWR for a potential new Finnish power plant. TVO awarded ARVEVA the contract for a new Finnish power plant, the 1,600 MWe EPR. [11, 12, 15, 16, 17, 20, 21, 22] Installations Under Construction or Planned Several 1,350 MWe ABWR units are under construction in Japan and Taiwan. Two ABWR s are nearing completion in Taiwan, and Japan has currently begun construction on its fourth ABWR (Shimane 3). Plant completion is planned for 2011, with major site work planned to start in In addition, three more ABWR s (Higashidori 1 and 2 and Ohma) are listed as on order by the Japanese utilities, with completion dates of 2012 or later. The ABWR is under consideration by South Texas Project Nuclear Operating Company (STPNOC)/NRG South Texas LP at the existing STP nuclear plant on its Bay City, TX site and by TXU at its Comanche Peak site. An Early Site Permit application for the Bay City site is planned to be submitted in September All of the above organizations will be applying for COLs during

39 The Nuclear Regulatory Commission staff received a letter of intent dated March 13, 2006 from Amarillo Power announcing their intent to pursue two GE Advanced Boiling Water Reactors (ABWRs). The applicant intends to submit an Early Site Permit (ESP) application before the last quarter of 2007, and a Combined Operating License (COL) application as soon thereafter as practicable. The March 13, 2006, letter contained proprietary information submitted under 10 CFR 2.390, which was subsequently made public through a July 27, 2006, letter from Amarillo Power. [11, 15, 16, 17, 20, 21] U.S. Design Differences The ABWR nuclear plant was developed in cooperation with the Tokyo Electric Power Company and GE's partners Hitachi and Toshiba. This effort culminated in the construction of the Kashiwazaki ABWRs, Units 6 & 7. At the same time, the ABWR was being reviewed and approved by the U.S. Nuclear Regulatory Commission. The ABWR Design Certification was issued and concurrently the First of a Kind Engineering (FOAKE) program, sponsored by DOE and 16 U.S. utilities through EPRI and cost-shared by DOE was completed. These efforts formed the basis for the Lungmen ABWRs. The Taiwan Power Company selected the ABWR from among several competing designs. While the ABWR design is usually associated in the U.S. with General Electric, variations on the design have also been built by Toshiba and Hitachi. [2, 11, 22] 3.2 General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) With Passive Safety Systems The economic simplified boiling water reactor (ESBWR) is a large scale boiling water reactor that uses economies of scale, proven technology, and components from the advanced boiling water reactor (ABWR) to create this new reactor type. [23] Plant Description General Electric Company s (GE) ESBWR is a 4,500 MWt (appropriately 1,600 MWe nominal) reactor design that uses natural circulation for normal operation and has passive safety features. Figure 3-2 shows the ESBWR safety system configuration. The ESBWR has a low-leakage containment vessel, which comprises the drywell and wetwell. The containment vessel is a cylindrical steel-lined reinforced concrete structure integrated with the reactor building. The containment boundary is illustrated as a dashed line on Figure 3-2, which also shows key features of the safety system configuration. The ESBWR Safety Systems design incorporates four redundant and independent divisions of the passive Gravity Driven Core Cooling System (GDCS), the Automatic Depressurization System (ADS) and a Passive Containment Cooling System (PCCS). Refer to Figure 3-2. Heat removal and inventory addition are also provided by the Isolation Condenser System (ICS) and the Standby Liquid Control System (SLCS). 3-11

40 The ADS function of the nuclear boiler system depressurizes the reactor pressure vessel (RPV) in sufficient time for the GDCS injection flow to replenish core coolant to maintain core temperature below design limits in the event of a LOCA (loss of coolant accident). The ADS also maintains the reactor depressurized for continued operation of GDCS after an accident without need for power. The ADS consists of safety relief valves (SRVs) and depressurization valves (DPVs) and their associated instrumentation and controls. The ICS removes decay heat after any reactor isolation during power operations. Decay heat removal limits further pressure rise and keeps the RPV pressure below the safety relief valves pressure setpoint. The ICS consists of four independent loops, each containing a heat exchanger that condenses steam on the tube side and transfers heat by heating/evaporating water in the Isolation Condenser/Passive Containment Cooling (IC/PCC) pools, which are vented to the atmosphere. The ICS passively removes sensible and core decay heat from the reactor (i.e., heat transfer from the IC tubes to the surrounding IC/PCC pool water is accomplished by natural convection, and no forced circulation equipment is required) when the normal heat removal system is unavailable. The PCCS maintains the containment within its pressure limits for design basis accidents such as a LOCA. The system is passive, and after initiation, no components move. The PCCS consists of six low pressure, totally independent loops, each containing a steam condenser (passive containment cooling condenser) that condenses steam on the tube side and transfers heat to water in a large cooling pool (IC/PCC pool), which is vented to the atmosphere. The ESBWR Standard Plant includes six (6) main buildings dedicated exclusively or primarily to housing systems and equipment related to a nuclear system or controlled access to these systems and equipment. These buildings include the reactor building (RB), the control building, the fuel building, the turbine building, the radwaste building and the electrical building. Figure 3-3 gives a typical power block arrangement for an ESBWR. Buildings (structures) and systems not in the ESBWR Standard Plant design include the main transformer; switchyard; heat sinks for the main condenser, decay heat, and system waste heat; sewage and water treatment building; and storage tanks for fuel oil, nitrogen and demineralized water. The Turbine Building (TB) encloses the turbine-generator, main condenser, condensate and feedwater systems, condensate purification system, offgas system, turbine-generator support systems, bridge crane, and various RB and TB auxiliary systems. Shielding is provided for the turbine on the operating deck. The turbine-generator and condenser are supported on a pedestal type foundation. The ESBWR Standard Plant has a containment structure comprised of a drywell and wetwell. The containment structure is a reinforced right circular cylindrical concrete vessel integrated with the reactor building. The standard plant uses a single-cycle, natural circulation BWR nuclear steam supply system (NSSS), designed by GE. The reactor pressure vessel (RPV) is a vertical, cylindrical pressure vessel of welded steel construction, with a removable top head. Four (4) main steam lines (MSLs) transport the steam generated in the reactor core from the RPV to the main turbine. Each MSL is connected to an outlet nozzle in the RPV and these MSLs contain two MSIVs (main steam lines isolation valves) in series. The inside of the main steam outlet nozzle, which is part of the RPV, has the shape of a 3-12

41 venturi type flow limiter. This MSL flow restrictor limits the coolant blowdown rate from the RPV in the event a MSL break occurs downstream of the nozzle. The flow restrictors also contain instrument line taps used for detecting and monitoring steam flow. Natural circulation of coolant water is provided in the ESBWR reactor design. This natural circulation in the ESBWR reactor is established by the density differences between the water in the reactor vessel annulus (outside the shroud and chimney) and the steam/water mixture inside the shroud and chimney. The colder higher density water in the annulus creates a higher pressure or a driving head when compared to the hotter lower density fluid (steam/water) in the core and chimney. The energy produced in the reactor core heats coolant and begins to convert the coolant water entering at the bottom of the core, into a steam/water mixture. The coolant water in the reactor core is first heated to the saturation temperature and then additional heat is added to the coolant and the boiling process of the core coolant is started. As the coolant steam/water mixture travels upward through the reactor core the percent of saturated steam increases until it exits the core. This steam/water mixture travels upward through the chimney to the steam separators where centrifugal force separates the steam from the water. The separated, saturated water returns to the volume around the separators while the slightly wet steam travels upward through the steam dryers and eventually out the main steam nozzle and piping to the turbine. The feedwater enters the reactor vessel at the top of the annulus and mixes with the saturated water around the separators and subcools this coolant water. The resulting mixture is subcooled below the saturation temperature and the cooler mixture then travels downward through the annulus to reenter the core. The coolant water therefore forms a recirculation loop within the reactor vessel. The mass of steam leaving the vessel is matched by the mass of feedwater entering the reactor vessel. Figure 3-4 shows the ESBWR reactor pressure vessel system key features and Figure 3-5 shows the ESBWR reactor pressure vessel natural circulation process. The proposed main turbine for the ESBWR has one high pressure (HP) turbine and three low pressure (LP) turbines. However, other turbine configurations could be selected by the COL applicant for use with the ESBWR. The steam passes through a moisture separator reheater (MSR) prior to entering the LP turbines. The steam exhausted from the LP turbines is condensed and degassed in the condenser. Steam is bled off from each turbine and is used to heat the feedwater. The turbine-generator (TG) system is nonsafety-related and is not needed to effect or support a safe shutdown of the reactor. The turbine generator is orientated within the turbine building to be inline with the reactor building to minimize the potential for any high energy TG system generated missiles from damaging any safety-related equipment or structures. The Main Condenser (MC) condenses the exhaust steam from the main turbine, provides a heat sink for the Turbine Bypass System (TBS), and is a collection point for other steam cycle drains and vents. The Circulating Water System (CWS) provides cooling water for removal of the power cycle waste heat from the main condensers and transfers this heat to the power cycle heat sink. The CWS has no safety design basis. The ESBWR has a design life of 60 years. [24, 25, 26] 3-13

42 Figure 3-2 ESBWR Safety System Configuration (not to scale) [25] 3-14

43 Figure 3-3 Typical ESBWR Power Block Arrangement 26] Legend: CB Control Building EB/TSC Electrical Building/Technical Support Center FB Fuel Building RB Reactor Building RW Radwaste Building TB Turbine Building WD Wash Down Bays (Equipment Entry) 3-15

44 Figure 3-4 ESBWR Reactor Pressure Vessel System Key Features [24] 3-16

45 Figure 3-5 ESBWR Reactor Pressure Vessel Natural Circulation Process [24] 3-17

46 3.2.2 Design Improvements of the ESBWR The ESBWR simplified design provides improved safety; excellent economics; better plant security; a broad seismic design envelope; and operational flexibility that increases plant availability. The following is a partial listing of the design improvements for the ESBWR: 1. Natural circulation of coolant water is provided in the ESBWR reactor design, enhancing plant performance and simplifying the plant design. Natural circulation is established because of density differences between water in the vessel annulus and the steam/water mixture inside the shroud and chimney. Natural circulation is enhanced by the shorter fuel, 8.6 meter (~28.2 feet) chimney, improved steam separator, and opening the flow path between the downcomer and the lower plenum. 2. Elimination of the recirculation pumps. 3. The ESBWR has fewer moving parts (pumps, valves, etc.) than the current BWR design. The ESBWR has achieved its basic plant simplification by incorporating innovative adaptations of operating plant systems into the plant design (i.e., combining shutdown cooling and reactor water cleanup systems). 4. Passive safety features eliminate the need for safety-grade pumps and ac power. 5. Design simplification also results in a reduction in building volume compared with the ABWR, even though generator output is increased by nearly 15 percent. 6. Larger electric output (1,600 MWe nominal). 7. Nearly all safety systems are located in the containment or directly above the building. See Figure 3-2. [24, 25 26, 27] Estimated Cost The overnight construction cost for an ESBWR has been estimated to be in the range of $1,160 to $1,250 per kwe; however, in the September 2006 issue of Power Engineering, a GE Nuclear Energy manager is quoted as stating that the ESBWR can be built for $1,350 per kwe. Thus, overnight construction costs estimates are increasing over time in the published data. Assuming a $1,350/kWe overnight cost for a 1,600 MWe unit would mean the unit would have an overnight construction cost of $2,160 million. These costs could grow to approximately $1,925/kWe once the owners cost and the AFUDC are included (approximately $3,080 million), see Table 3-4 for cost estimates (Note: These cost estimates were not directly provided by the reactor vendor and are not endorsed by the reactor vendor. The cost estimates were developed based on the information that was publicly available. No guarantee is provided as to the accuracy of these values. For example, these values would change as a result of increasing material, equipment and labor prices). The December 7, 2006 issue of Platts Nucleonics Week stated that earlier this year (2006), executives at Westinghouse, AREVA NP and General Electric estimated the cost for a new reactor at between $1,600 and $2,000 per installed kilowatt. [13, 23, 25, 28] 3-18

47 Table 3-4 Technology Basis Data (ESBWR) Plant Type General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) With Passive Safety Systems Fuel Type Fuel Cycle Enriched Uranium Dioxide 4.2% GE14 (10x10 fuel assembly) [23] 24 month cycle [23] 25 day outage duration [6] Number of Units 1 [6] Unit Size, MW 1,600 MWe (gross) at 4,500 MWt, electrical output values can vary as much as ± 50 MWe depending on the Turbine Island design and sitespecific conditions. (To be confirmed in COL phase) [25] Plant Size, Net MWe 1,535 Plant Auxiliaries, MWe 65 [6] Plant Capacity Net 96% Available for Commercial Orders, Year Now First Commercial Service (Technology Year) 2014 Hypothetical In-Service Year 2014/2015 in the U.S. [27] Design & Cost Estimate Rating Technology Development Rating Plant Location E - Goal E Laboratory U.S. Plant Capital Cost Structures & Improvements Reactor Plant Equipment Turbine Plant Equipment Electrical Plant Equipment 1 Unit $250 million ($182/kWe) $450 million ($328/kWe) $300 million ($219/kWe) $120 million ($88/kWe) $50 million ($36/kWe)for miscellaneous plant equipment Other BOP Facilities $20 million ($15/kWe) for the Main Condenser Heat Rejection System $160 million ($117/kWe)for construction services, engineering / home office services and field supervision / field office services 3-19

48 Table 3-4 Technology Basis Data (ESBWR) (continued) Plant Type General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) With Passive Safety Systems Installation General Facilities & Engineering Fee Project & Process Contingency Total Plant Cost Total Cash Expended (Mixed year $) AFUDC (interest during construction) Total Plant Investment (includes AFUDC) Total Owner Costs Total Capital Requirement, Hypothetical In-Service Year (includes AFUDC) $450 million ($328/kWe) manual labor $360 million ($263/kWe) non manual labor See Other BOP Facilities above $260 million ($190/kWe) (12% of the plant capital cost of $2,160 million) $2,420 million ($1,765/kWe) To be supplied by EPRI $380 million ($277/kWe) (Based on 8% interest, 54 million per month for 45 months) $2,800 million ($2,042/kWe) $280 million ($204/kWe) (Based on 10% of the total plant cost) To be supplied by EPRI Cost for Hypothetical In-Service Year Operating & Maintenance Costs Fixed $/MW-yr [6] (staff of 701 at a Greenfield site) Incremental, mils/kwh: 4.64 fuel cost Variable (includes consumables), mills/kwh 2.45 variable 1 Nuclear Waste Fee 8.09 total [10] Net Plant Heat Rate, Btu/kWh (if applicable) Full Load Min Load Average Annual Information not available Information not available Information not available 3-20

49 Table 3-4 Technology Basis Data (ESBWR) (continued) Plant Type General Electric Advanced Economic Simplified Nuclear BWR (ESBWR) With Passive Safety Systems Unit Availability Equivalent Planned Outage Rate, % 3.5 Equivalent Unplanned Outage Rate, % 4 Forced Outage Rate [6] Equivalent Availability, % 92.5% Capability Ratio 95 Capacity Factor [6] Duty Cycle Minimum Load, % Information not available Preconstruction, License & Design Time, Years 4 to 5 Idealized Plant Construction Time, Years 3.25 (39 months) [6] Unit Life, Years (minimum) Estimated Land Area Needed (acres) Excluding cooling ponds The EAB would be about 0.5 miles from the containment. Estimated Water Requirements (gpm) 41, Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. 2. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline Deployment Plans Summary ESBWR is the latest in a long line of proven GE BWR reactors. ESBWR employs passive safety design features. It is a simplified reactor design, allowing faster construction and lower costs. A GE-designed Generation III+ reactor, ESBWR is currently in the U.S. Design Certification process. An initial Safety Evaluation Report (SER) is expected in 2007, in time for Construction and Operating License (COL) submissions that would support the commercial operation of new ESBWRs by GE is ready to support utilities looking to build an ESBWR nuclear power plant, with a well established global supply chain. 3-21

50 Table 3-5 Technology Monitoring Guide (ESBWR) Leading Developers of the Science or Technology Unresolved Issues Changes To Watch For Major Trends Leading Vendors Industrial Firms Nonprofit Organizations Government Organizations R&D Intensity Technologies ABWR The development program for the ESBWR was directed to design a larger reactor that used economies of scale, proven technology, and components from the ABWR to create a new reactor at reduced capital cost. There is high confidence in the design because it uses standard, proven equipment including extensive use of ABWR components and fully tested passive safety features [23] DOE EPRI including the BWR Owners Group GE GE [2] Increase in MWe output First plant order in the U.S. Design certification by the U.S. NRC. Completion of detailed design. 3-22

51 Table 3-6 Technology Development and Assessment (ESBWR) Features and Characteristics of Technology Major Technical Issues Key Vendor Activities Resource Requirements That Impact Technology Key Business and Market Indicators Current Status Simplification, standardized design, operational flexibility, improved economics Natural circulation of coolant water is provided in reactor design [27] Detailed design is now being completed COL Activities in the U.S. Design Finalization None COL applications to the NRC that use the ESBWR as the reference plant Future Considerations or Trends Information not available Information not available Information not available Information not available Start of construction of an ESBWR in the U.S. Key technology needs Information not available Information not available Technology Outlook Anticipate several Combined License (COL) applications, by numerous utilities, will be submitted in 2007/2008. Information not available Development Timeframe Research - early 1990 s [27] Information not available Research Ongoing Information not available Development Ongoing Information not available Demonstration No demonstration plant needed, the ESBWR is based in proven BWR technology Information not available Projected Commercialization Date 2014/ U.S. [27] Information not available The technology evaluation was carried out by the utilities, aided by EPRI and its contractors. GE submitted the GE ESBWR Design Control Document (DCD) application to the NRC. The NRC formally accepted the GE ESBWR DCD application via a letter dated December 1, 2005 and established a schedule for its review, with a milestone of October 11, 2007, for issuance of the safety evaluation report (SER) with open items. This DCD is still under review by the NRC. [20, 29] 3-23

52 3.2.5 Installations Under Construction or Planned No ESBWR are currently under construction. A consortium led by Dominion Resources (Dominion) submitted a proposal to the DOE (U.S. Department of Energy), in March 2004, to demonstrate the NRC s process for licensing the construction and operation of new nuclear power plants. The Dominion consortium has selected the GE ESBWR technology. The proposed site location is at the North Anna site in Virginia and up to two units may be proposed. On April 26, 2004, NuStart submitted its proposal to DOE to demonstrate NRC s process for licensing the construction and operation of new nuclear power plants. On May 6, 2005, NuStart issued a press release stating that it had signed a cost-sharing agreement with DOE. In a letter dated November 17, 2005 to the NRC, NuStart announced the Grand Gulf site will be a singleunit site and will reference the GE ESBWR design and the Grand Gulf ESP, if granted. Entergy plans to submit a COL application for the GE ESBWR design in early The planned site location is at the River Bend site in Louisiana. A single unit ESBWR plant is planned. [20, 30] U.S. Design Differences None, the GE ESBWR design was developed for construction in the U.S. 3.3 Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) Canada s CANDU (Canadian Deuterium) reactor designs use multiple pressure tubes containing nuclear fuel assemblies in the active core region, which permit on-line refueling. Heavy water is pumped through the pressure tubes to remove heat and is also used to moderate neutrons in a low pressure vessel (the Calandria) that surrounds the pressure tube region. CANDU reactors have been deployed outside Canada (e.g., China, Romania, South Korea and Argentina). Recent advances to this design use light water cooling but retain heavy water moderation in the Calandria. This approach holds significant promise for improved maintainability and economics. Most CANDU designs are in the medium (500-1,000 MWe) size range. Atomic Energy of Canada Limited (AECL) has developed the ACR-700 (Advanced CANDU Reactor-700) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor, low operating cost, increased operating life, simple component replacement and enhanced safety features. The ACR-700 is 750 MWe but is physically much smaller, simpler and more efficient as well as 40% cheaper than the CANDU 6. But the ACR-1000 (approximately 1,200 MWe) is now the focus of attention by AECL. The ACR-1000 has more fuel channels than the ACR-700 (292 fuel channels vs. 520 for the ACR-1000). Each fuel channel can be regarded as a module of approximately 2.3 MWe. The ACR-700 and ACR-1000 are light water cooled, heavy water 3-24

53 moderated pressure-tube reactors as opposed to using heavy water for both cooling and moderation as in the original CANDU design. It should be noted that the discussion that follows focuses on the ACR-1000 design. [11, 31, 32] Plant Description The Advanced CANDU Reactor (ACR) technology for the ACR-1000 is currently in the Canadian pre-licensing process that builds on the CANDU 6 experience to deliver improved safety margins, lower capital and operating costs, improved maintenance and excellent operating performance with high availability, long life, and fast construction. The CANDU 6 nuclear power reactor design is an AECL proven design, with operating reactors in five different countries and an excellent safety and reliability record. The ACR design is based on the use of modular horizontal fuel channels surrounded by a heavy water moderator, this is the same design that is used in all CANDU reactors. The major innovation in ACR is the use of low enriched uranium fuel and light water as the coolant, which circulates in the fuel channels. This results in a more compact reactor design and a reduction of heavy water inventory, both contributing to a significant decrease in cost compared to the current CANDU reactor designs that employ natural uranium as fuel and heavy water as coolant. Figure 3-6 is an overall ACR-1000 plant simplified flow diagram. Figure 3-6 Overall ACR-1000 Plant Simplified Flow Diagram (Provided by AECL) [32] 3-25

54 A single unit ACR would have the following structures: Reactor Building Reactor Auxiliary Building Turbine Building Main Control Building Secondary Control Area Maintenance and Service Buildings Condenser Cooling Water Pumphouse Raw Service Water Pumphouse Main Switchyard The major nuclear systems of an ACR-1000 plant are located in the reactor building (RB) and reactor auxiliary building (RAB). These buildings are robust and shielded where necessary to control radiation dose. The RB is a pre-stressed concrete structure that is designed to be seismically qualified and tornado proofed and is the principal component of the containment system. The concrete outer walls of the RB have an inner steel liner. The RAB is also designed to be seismically qualified and tornado proofed. The RAB is a reinforced concrete and steel structure that surrounds the RB and contains the long-term cooling pumps and heat exchangers, the spent fuel bay cooling and purification system pumps and heat exchangers, the re-circulated cooling water pumps, heat exchangers, and valve stations. The safety and isolation valves for the main steam lines are located in a seismically qualified concrete structure that is on top of the RAB. Each corner of the reactor auxiliary building houses redundant safety equipment in a four-quadrant configuration. The turbine building will be located to one side of the RAB, so that the turbine shaft is aligned perpendicular to the RB. The turbine building contains the turbine generator, the condenser, the condensate and feedwater systems, the building heating plant, and any compressed gas required for the balance of plant. The turbine building is equipped with blow-out panels in the walls and roof to relieve internal pressure in the event of a steam line rupture. The main control building is seismically-qualified and tornado protected and is constructed from a superstructure of steel and reinforced concrete and reinforced concrete substructure. The building contains the main control room and electrical equipment. The secondary control area is completely separate for the main control building and has sufficient control and monitoring equipment to shutdown the unit, initiate required cooling and ensure a safe, maintained shutdown state should the main control room become unavailable. The seismically-qualified maintenance and service buildings contain all the conventional plant services, including radioactive waste handling facilities, heavy water management systems, equipment storage, maintenance shops, and locker rooms for the staff. The building will have a reinforced concrete substructure with a steel framed superstructure. The condenser cooling water pumphouse will have a reinforced concrete substructure and a braced steel framed superstructure. The structure will contain the condenser cooling water pumps, non-safety related raw service water pumps, screen wash pumps, trash racks, screens, 3-26

55 and chlorination equipment if required. The raw service water pumphouse is seismicallyqualified and tornado protected and is constructed from a braced steel frame superstructure and a reinforced concrete substructure. The plant layout is also designed to achieve the shortest practical construction schedule while supporting easier maintenance practices. Buildings are arranged to minimize interferences during construction, with allowance for on-site fabrication of module assemblies. Through the use of open-top construction, provisions exist for flexible equipment installation sequences. The footprint of the two-unit plant has been minimized with the adoption of common areas for the main control room, service and maintenance buildings. A single-unit plant can be adapted from the two-unit layout with no significant changes to the reference design. The reactor assembly is designed to use LEU (low enriched uranium - Average <2.0 wt% 235 U) fuel and light water coolant. The use of LEU fuel enables the ACR-1000 designs to operate with a negative void reactivity. Similar to the CANDU 6 design, the ACR-1000 design has an efficient, low pressure heavy water moderator and uses low neutron absorbing zirconium alloys for the core structures, including horizontal fuel channels and fuel cladding that contains the fuel. The design of the ACR-1000 has small diameter horizontal fuel channels that contain high pressure, high temperature heat transport system coolant and allows the use of a separate low pressure moderator system in which the reactivity control devices operate. Fuel is replaced while the reactor is at power, in order to maintain sufficient positive reactivity. This feature contributes to high availability factors and improved outage flexibility since refueling outages at fixed cycle times are not required. For the ACR-1000 there are 12 (43-element) CANFLEX -ACR fuel bundles in each of the 520 fuel channels. The center element of the fuel bundle contains neutron absorbers, while the remaining elements contain U-235 enriched UO 2 pellets. A burnable absorber is used in some of the elements that contain enriched pellets to optimize the power rating of the fuel. The neutron absorbers of the center element are used for management of coolant void reactivity. A very thin layer of graphite (called CANLUB) covers the inside surface of the fuel cladding to enhance fuel performance. The CANLUB between the pellet and sheath reduces friction and increases stress corrosion cracking limits. Figure 3-7 shows a CANFLEX - ACR fuel bundle. The ACR-1000 inherent feature for operating with neutron absorbers makes it ideally suited to burn other fuel types such as mixed oxides (MOX) and thorium. 3-27

56 Figure 3-7 CANFLEX -ACR Fuel Bundle (Provided by AECL) [32] The reactor power distribution is controlled by the fuel handling system which supplies on demand refueling of the reactor. The fuel handling system for the ACR design is developed from the CANDU 6 design. The current CANDU 6 design has been improved to increase overall performance of the fuel handling and storage system. The improvements include replacing all hydraulic drives with electric drives and the elimination of heavy water in the fuel handling system, since the reactor coolant system uses light water. The fuel handling system stores and handles fuel, from the arrival of new fuel to the storage of spent fuel. The fuel handling system consists of two fueling machine heads each mounted on a fueling machine bridge and columns that are located at each end of the reactor. Spent fuel baskets are stacked in frames under water in the spent fuel storage bay. The storage bay has a capacity of at least 10 years of operation and provisions are provided in the design for direct transfer to dry fuel storage. At power refueling assures that very little reactivity is needed in movable control rods or in poisons dissolved in the moderator (no chemicals are added to the reactor coolant for reactivity control). Therefore, any malfunctions in the reactor control system produces only small reactivity changes in the reactor. The main steam supply system has higher pressure and temperature conditions than the current CANDU designs. This higher pressure and temperature improves the turbine cycle efficiency. The heat transport system (HTS) has two-loops, each single-loop is in a figure eight configuration with two steam generators, two heat transport pumps (11.5 MWe each), two reactor outlet headers and two reactor inlet headers. The pressurized light water coolant is circulated through the reactor fuel channels (520 channels in the ACR-1000 design) to remove the heat produced by nuclear fission in the core. When the reactor is at power the pressurizer accommodates changes in the reactor coolant volume from zero power to full power; thus, allowing rapid power changes without tasking the coolant feed and bleed system. Reactor pressure is controlled in the pressurizer by electrical heaters that add heat to increase pressure 3-28

57 and by cold water sprays that remove heat via the reactor inlet heaters to reduce the pressure of the system. Reactor pressure is also controlled by the feed and bleed system when the pressurizer is isolated at low reactor power or during heat-up and cool-down and when the reactor is shut down. Figure 3-8 shows the ACR-1000 nuclear system schematic. The heat that is removed by the coolant is used in the steam generators to produce steam. The ACR-1000 steam generator tubing is made from Incoloy-800. On the secondary side of the steam generators the steam is used to drive the turbine generator which produces electrical energy. Steam wetness at the steam nozzle has been reduced to 0.1%, based on latest steam separator technology, leading to improved turbine cycle economics. The ACR-1000 moderator system is a low-pressure, low-temperature system that is fully independent from the HTS. The moderator system consists of pumps and heat exchangers that circulate the heavy water moderator (D 2 O) through the reactor vessel and removes heat generated within the moderator during reactor operation. The heavy water acts as both moderator and reflector of the neutrons in the reactor core. The expected operating life of an ACR-1000 is 60 years taking into consideration the requirements of a mid-life outage for pressure tube replacement and major plant refurbishment activities. The fuel channels are designed to achieve a 30-year operating life with 90% capacity factor over the operating life of 60 years and a year-to-year expected capacity factor is 95%. [31, 32] 3-29

58 Figure 3-8 ACR-1000 Nuclear Systems Schematic (Provided by AECL) [32] 3-30

59 3.3.2 Design Improvements of the ACR-1000 The ACR-1000 design has several advantages over the parent CANDU design. 1. An enhanced safety feature of the ACR-1000 design includes a core design with a small negative coolant void reactivity and larger thermal margins than the current CANDU reactor designs. 2. Improved fuel burn-up through the use of low enriched uranium (LEU) fuel, contained in advanced CANFLEX -ACR fuel bundles. 3. The ACR-1000 safety systems are designed to mitigate the consequences of plant process failures, ensuring reactor shutdown, removal of decay heat and prevention of radioactive releases to the environment. 4. The ACR-1000 design incorporates two passive, fast acting, fully capable, diverse and separate shutdown systems that are physically and functionally independent of each other. Shutdown System 1 consists of mechanical shutoff rods that drop by gravity into the core when a trip signal de-energizes the clutches that hold the shutoff rods out of the core. The design of the shutoff rods is based on the proven CANDU 6 design. The in-core portion of the shutoff rods has been designed to accommodate the smaller ACR-1000 core lattice pitch. Shutdown System 2 injects a concentrated solution of gadolinium nitrate into the lowpressure moderator to quickly render the core sub-critical. The gadolinium nitrate solution is dispersed uniformly with pressurized gas, maximizing shutdown effectiveness. 5. Large concrete reactor vault, surrounding the core in the calandria vessel, contains a large volume of light water to further slow down or arrest severe core damage progression by providing a second, passive core heat sink. 6. A more compact core size due to a smaller fuel channel lattice pitch than used in the CANDU 6 design, thus reducing the heavy water requirements. 7. The use of light water as coolant as opposed to heavy water results in an over all reduction of the heavy water inventory for the plant and results in a reduction of systems needed for heavy water coolant cleanup and recovery and simplification of containment atmosphere cleanup systems. 8. Simplified reactor control through negative feedback in reactor power. 9. Flattened axial and radial profiles to optimize channel thermal power output. 10. The fuel handling system for the ACR design replaces all the hydraulic drives with electric drives. 11. The elimination of the use of heavy water in the fuel handling system is possible since the reactor coolant system uses light water. 12. Elevated reserve water tank (RWT) in upper level of the containment building delivers passive make-up cooling water by gravity to heat transport system, steam generators, moderator and the calandria vault. In addition, part of the reserve water tank inventory is used for containment sprays to remove heat from the containment atmosphere. This delays 3-31

60 progression of severe accidents and provides even more time for mitigating actions by the operator. A severe accident is one in which the fuel is not cooled within the heat transport system. ACR/CANDU design principle is to prevent severe accidents and to mitigate severe accident events, in addition to minimizing their consequences. Figure 3-9 shows the reserve water system that can supply water to the reactor and the steam generators (only one loop of the ACR-1000 heat transport system is shown). 13. Improved plant thermal efficiency through use of higher pressures and higher temperatures in the coolant and steam supply systems. 14. Approximately 30% reduction in spent fuel quantities compared to current operating CANDU plants. [32] 3-32

61 Figure 3-9 ACR-1000 Reserve Water System (Provided by AECL) [32] 3-33

62 3.3.3 Estimated Cost The December 7, 2006 issue of Platts Nucleonics Week stated that earlier this year (2006), executives at Westinghouse, AREVA NP and General Electric estimated the cost for a new reactor at between $1,600 and $2,000 per installed kilowatt. AECL indicated that the ACR-1000 is fully competitive in this cost range. Assuming a $2,000/kWe overnight cost for a 1,165 MWe unit would mean the unit would have an overnight construction cost of $2,330 million. These costs could grow to around $ 2,800/kWe once the owners cost and the AFUDC are included (~$3,270 million), see Table 3-7 for cost estimates (Note: These cost estimates were not directly provided by the reactor vendor and are not endorsed by the reactor vendor. The cost estimates were developed based on the information that was publicly available. No guarantee is provided as to the accuracy of these values. For example, these values would change as a result of increasing material, equipment and labor prices). [11, 13] Table 3-7 Technology Basis Data (ACR-1000) Plant Type Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) Fuel Type Slightly Enriched Uranium Dioxide - Average <2.0 wt% 235 U. 43-element CANFLEX -ACR fuel bundle design [6, 32] On line refueling Fuel Cycle Projected maintenance outage of 21 days every 3 years and a mid-life outage for pressure tube replacement and major plant refurbishment activities. This outage should be less than 1 year in length. [32] Number of Units 1 Unit Size, MW 3,180 MWth 1,165 MWe (gross) [32] Plant Size, Net MWe 1,085 [32] Plant Auxiliaries, MWe 80 total 11.5 MWe for each of the 4 heat transport pumps Plant Capacity Net 93% Available for Commercial Orders, Year Now First Commercial Service (Technology Year) 2016 Hypothetical In-Service Year

63 Table 3-7 Technology Basis Data (ACR-1000) (continued) Plant Type Design & Cost Estimate Rating Technology Development Rating Plant Location Plant Capital Cost Structures & Improvements Reactor Plant Equipment Turbine Plant Equipment Electrical Plant Equipment Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) E Goal E Laboratory Canada $270 million ($249/kWe) $480 million ($442/kWe) $320 million ($295/kWe) $120 million ($111/kWe) 1 Unit $50 million ($46/kWe) for miscellaneous plant equipment Other BOP Facilities $20 million ($18/kWe) for the Main Condenser Heat Rejection System $170 million ($157/kWe) for construction services, engineering / home office services and field supervision / field office services Installation General Facilities & Engineering Fee Project & Process Contingency Total Plant Cost Total Cash Expended (Mixed year $) AFUDC (interest during construction) Total Plant Investment (includes AFUDC) Total Owner Costs Total Capital Requirement, Hypothetical In-Service Year (includes AFUDC) $500 million ($461/kWe) manual labor $400 million ($369/kWe) Non manual labor See Other BOP Facilities above $280 million ($258/kWe) (12% of the plant capital cost of $2,330 million) $2,610 million ($2406/kWe) To be supplied by EPRI $360 million ($332/kWe) (Based on 8% interest, 62 million per month for 42 months) $2,970 million ($2737/kWe) $300 million ($276/kWe) (Based on 10% of the total plant cost) To be supplied by EPRI 3-35

64 Table 3-7 Technology Basis Data (ACR-1000) (continued) Plant Type Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) Cost for Hypothetical In-Service Year Operating & Maintenance Costs Fixed $/MW-yr $9.81 (Base on an ACR-700 greenfield twin unit site with 761 staff onsite and offsite) [6] Incremental, mils/kwh: 4.64 fuel cost Variable (includes consumables), mills/kwh 2.45 variable 1 Nuclear Waste Fee 8.09 total [10] Net Plant Heat Rate Full Load Min Load Average Annual 3180 MWth The Nuclear Steam Plant (NSP) does not impose load restrictions on power operation. Loadfollowing the grid provides up to 2.5% power variation, while operating at 97.5%. The design allows for rapid load reduction from steady statefull power 100% load to 75%, and as low as 50% when required (periodic load cycles). In addition, the plant is also capable of transition to continued operation at house load on loss of grid power, without subsequent turbine trip. N/A Unit Availability Equivalent Planned Outage Rate, % 2 (21 days every 36 months) [6] Equivalent Unplanned Outage Rate, % 4 (forced outage rate) [6] Equivalent Availability, % 94 [6] Capability Ratio 90% capacity factor over the operating life of 60 years and a year-to-year expected capacity factor of 95%. Duty Cycle Minimum Load, % The Nuclear Steam Plant (NSP) does not impose load restrictions on power operation. 3-36

65 Table 3-7 Technology Basis Data (ACR-1000) (continued) Plant Type Atomic Energy of Canada Limited (AECL) Advanced CANDU Reactor (ACR) Preconstruction, License & Design Time, Years 5 Idealized Plant Construction Time, Years 3.5 (42 months) 60 Unit Life, Years The fuel channels are designed to achieve a 30- year operating life and would be replaced once during the life of the unit. [31] 260 (minimum) Estimated Land Area Needed (acres) Excluding cooling ponds The EAB would be about 0.3 miles from the containment. Estimated Water Requirements (gpm) 28, Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. 2. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline Deployment Plans Summary The vendor s construction strategy is to use an open-top construction method using a very heavy lift crane. The vendor also plans to employ a construction strategy that provides concurrent construction, modularization and prefabrication and the use of advanced technologies to minimize interferences. The open-top/vertical installation construction method provides for an improved construction logic that reduces costs while reducing the schedule risk. The internal structure of the reactor building is initially built as vertical walls without floors. Major modules, including the floors, are then installed in parallel. The first ACR-1000 is expected to be built in Canada, and producing electricity by [32, 33] 3-37

66 Table 3-8 Technology Monitoring Guide (ACR) Leading Developers of the Science or Technology Unresolved Issues Changes To Watch For Major Trends Leading Vendors Industrial Firms Nonprofit Organizations Government Organizations R&D Intensity Technologies ACR-1000 On-going research (development and testing) CNSC 1 KAERI 2 3 COG AECL 4 AECL 4 Reduce the use of heavy water in the CANDU designs Use of low enriched uranium as fuel in the CANDU designs The development of a heavy water moderated design with supercritical light-water coolant. In a supercritical system the reactor operates above the critical point of water (22.4 MPa and 374 C) resulting in higher thermal efficiency than current LWR designs [3] Design certification by the U.S. NRC 1 Canadian Nuclear Safety Commission 2 Korea Atomic Energy Research Institute 3 CANDU Owners Group 4 Atomic Energy of Canada Limited 5 Light Water Reactor 3-38

67 Table 3-9 Technology Development and Assessment (ACR) Features and Characteristics of Technology Major Technical Issues Current Status Passive safety systems Modular construction Modular, horizontal fuel channel core Separate low temperature and pressure moderator Reactor vault filled with light water surrounding the core At power refueling Reactor building access for at power maintenance Detailed design is now being completed Future Considerations or Trends The development of a heavy water moderated design with supercritical light-water coolant. In a supercritical system the reactor operates above the critical point of water (22.4 MPa and 374 C) resulting in higher thermal efficiency than current LWR designs Information not available Key Vendor Activities Design Finalization Information not available Resource Requirements That Impact Technology Key Business and Market Indicators Heavy water available Plant orders Canada regulatory approval Information not available Start of construction of a ACR Key technology needs Information not available Information not available Technology Outlook Development Timeframe First plants will be built in Canada Information not available Research Ongoing Information not available Development Ongoing Information not available Demonstration No demonstration plant needed, the ACR-1000 is based on proven CANDU technology Information not available Projected Commercialization Date 2016 Information not available 3-39

68 Currently, there are no immediate plans for licensing the ACR-1000 design in the U.S. On June 19, 2002, Atomic Energy of Canada, Limited (AECL) requested pre-application review of their ACR-700 for licensing in the United States. The CNSC (Canadian Nuclear Safety Commission) has not licensed a new nuclear generating station since In preparation for the possibility of licensing new nuclear generating stations, in the CNSC commenced the production of a Licensing Basis (LB) document that will be used to assess the licensability of new reactors in Canada. The LB document will be applied to the Advanced CANDU Reactor (ACR) being designed by Atomic Energy of Canada Limited and to any other proposed reactor designs. [34] Installations Under Construction or Planned The CANDU 6 reactor has had units built recently in China and a construction of a second unit is near completion in Romania. The first ACR-1000 unit is expected to be operating in 2016 in Ontario, Canada. Currently, there are no potential combined license applications for the ACR-700 or ACR-1000 reactors scheduled with the U.S. NRC. [10, 20, 33] U.S. Design Differences The ACR-1000 reference plant design and configuration have been developed taking into consideration the U.S. Regulatory Requirements established in 10 CFR 50, in order to ensure a readily adaptable generic design. System configuration and design features, including layout, are part of the reference basis; thus, they will remain unchanged and have been designed to meet the regulatory requirements in the U.S. Only details such as selection of equipment, component design or analyses required to demonstrate specific NRC acceptance criteria are different for the U.S. approach than those employed to satisfy the Canadian regulatory requirements. [32] 3.4 PBMR (Pty) Ltd. (South Africa), Pebble Bed Modular Reactor (PBMR) The PBMR (Pebble Bed Modular Reactor) is an advanced helium-cooled, graphite-moderated high temperature gas-cooled reactor with a direct closed-cycle, gas turbine power conversion system. A 400 MWt (165 MWe nominal) Demonstration Power Plant (DPP) for the production of electricity is being developed in South Africa for its national utility Eskom. Although, it is not the only High Temperature Reactor (HTR) currently being developed in the world, the South African project is internationally regarded as having the most advanced technology in the field. The PBMR design provides very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. Billiard ball-sized pebbles of ceramic graphite, impregnated with thousands of tiny, coated particles of lowenriched uranium, fuel the PBMR in a continuous process that eliminates refueling outages. Online refueling which increases availability and load following. Operators feed fresh pebbles into the top of the reactor. The pebbles then travel through the core, coming to rest at the bottom. 3-40

69 For optimal fuel use, pebbles travel through the core six times before being discharged to the used-fuel tanks. Since the uranium and subsequent used fuel products are embedded in the graphite fuel spheres, separation of material is more difficult, increasing the proliferation resistance of the design. Due to its high operating temperature of 900 C (1,652 F), the PBMR also offers promising process heat applications. Applications identified so far include steam-methane-reforming for hydrogen, high quality steam for oil sands extraction, bulk hydrogen production, co-generation for petrochemical industries and desalination. These applications offer significant commercial promise with present day natural gas prices and growing carbon constraints and are being considered as possible demonstration projects with potential industrial customers. Substantial experience in PBMR technology was gained by the development of the technology in Germany. Two pebble bed reactors where constructed and operated in Germany, the AVR and THTR. A third one, the HTR Module, was partially designed but was not constructed. China recently constructed and is now operating a small test reactor, the HTR 10. In 2003, the PBMR design was reviewed and included as one of the potential Generation IV International Forum reactor systems for early very high temperature gas reactor deployment. In parallel with the Generation IV program, the U.S. DOE has developed plans for the Next Generation Nuclear Plant (NGNP) for Hydrogen and Electricity. The program was authorized in the Energy Policy Act of In 2006, a Westinghouse-led, PBMR-based consortium was competitively selected for the first stage NGNP engineering work. This report will focus on the pebble bed reactor technologies being utilized in the development of the South Africa design. [35] Plant Description The PBMR has a vertical steel reactor pressure vessel, which has a 6.2 m (~20.3 ft.) inner diameter, and is about 27 m (~88.6 ft.) high. The reactor pressure vessel contains and supports a metallic core barrel, which in turn supports the annular pebble fuel core. This annular fuel core is located in the space between central and outer graphite reflectors. Vertical borings in these reflectors are provided for the reactivity control elements. Two diverse reactivity control systems are provided for shutting the reactor down independently; one being control rods in the outer reflector, and the other being small absorber spheres which are dropped into borings in the central reflector. A schematic diagram and the physical layout of the main power system are shown in Figure 3-10 and Figure 3-11, respectively. 3-41

70 Figure 3-10 Simplified Schematic Diagram of the PBMR Main Power System (Provided by PBMR (Pty) Ltd) [35] 3-42

71 Recuperator Compressor Pre-cooler Power Turbine Maintenance Shut-off Valve (x2) Gearbox Core Conditioning System Generator Core Barrel Conditioning System Intercooler Maintenance Shut-off Valve Figure 3-11 Physical Layout of the PBMR Main Power System (Provided by PBMR (Pty) Ltd) [35] The PBMR uses particles of enriched uranium dioxide coated with silicon carbide and pyrolitic carbon. The particles are encased in graphite to form a fuel sphere or pebble about the size of a billiard ball. Helium is used as the coolant and energy transfer medium, to drive a closed cycle gas turbine-compressor and generator system. When fully loaded, the core would contain approximately 452,000 fuel spheres. To remove the heat generated by the nuclear fission reaction, helium coolant enters the reactor vessel at a temperature of 500 C (932 F) and a pressure of 9 MPa (1,305 psi). The gas flows down between the hot fuel spheres (pebbles), after which it leaves the bottom of the vessel, having been heated to a temperature of 900 C (1,652 F). The hot gas then enters the turbine, which is mechanically connected to the generator through a speed-reduction gearbox on one side, and to the gas compressors on the other side. The coolant leaves the turbine at 500 C (932 F) and 3.0 MPa (435 psi), after which it is cooled, recompressed, reheated and returned to the reactor core. The thermodynamic cycle used is a Brayton cycle with a water-cooled intercooler and pre-cooler. A high-efficiency recuperator is used after the power turbine. The helium, cooled in the recuperator, is passed through the pre-cooler, low-pressure compressor, the intercooler and high-pressure compressor before being returned through the recuperator to the reactor core. 3-43

72 The pre- and intercoolers are cooled on the secondary side by water in a closed circuit. This closed circuit, in turn, is cooled by the seawater through a secondary heat exchanger. The Demonstration Power Plant is being designed to be cooled with seawater; however, fresh water could also be used as the coolant. The power taken up by the helium in the core and the power given off in the power turbine is proportional to the helium mass flow rate for the same temperatures in the system. The mass flow rate depends on the pressure, so the power can be adjusted by changing the pressure in the system. The operation of the reactor at high pressure and high temperature results in a relatively high thermal efficiency. While a typical light water reactor has a thermal efficiency (electrical power output divided by core thermal power) of approximately 33%, an efficiency of about 41% is anticipated in the basic PBMR design. Online refueling is another key feature of the PBMR. The fuel is introduced at the top of the reactor while used fuel is removed at the bottom to keep the reactor at full power. Figure 3-12 is a schematic diagram of the fuel handling system during normal operation. Figure 3-13 shows an image of the physical layout of the fuel handling and storage system. 3-44

73 Figure 3-12 Schematic Diagram of the PBMR Fuel Handling System During Normal Operation (Provided by PBMR (Pty) Ltd) [35] 3-45

74 Figure 3-13 Physical Layout of the PBMR Fuel Handling and Storage System (Provided by PBMR (Pty) Ltd) [35] The aim is to operate uninterrupted for six years before the reactor is shut down for scheduled maintenance. However, for the demonstration module, a number of interim shutdowns will be required for planned evaluation of component and system performance. Shutdown will be done by inserting the control rods. Start-up is affected by making the reactor critical through withdrawal of control rods. Then, by using the nuclear heat generated in the core and initially motoring the turbine generator system to provide gas circulation, the Brayton cycle will initiate and become self-sustaining. How the PBMR Fuel Works PBMR fuel is based on a proven, high-quality German fuel design consisting of low enriched uranium triple-coated isotropic (LEU-TRISO) particles contained in a molded graphite sphere. A coated particle consists of a kernel of uranium dioxide surrounded by four coating layers as shown in Figure

75 Figure 3-14 PBMR Fuel Element System (Provided by PBMR (Pty) Ltd) [35] In the fabrication process, a solution of uranyl nitrate is dropped from small nozzles to form micro-spheres, which are then gelled and calcined (baked at high temperature) to produce uranium dioxide fuel kernels. The kernels are then processed in a Chemical Vapour Deposition (CVD) furnace in which the different coating layers are added. For PBMR fuel, the first layer deposited on the kernels is porous carbon. This is followed by a thin coating of pyrolytic carbon (a very dense form of carbon), a layer of silicon carbide (a strong refractory material), and another layer of pyrolytic carbon. The porous carbon accommodates any mechanical deformation that the uranium dioxide kernel may undergo during the lifetime of the fuel, as well as gaseous fission products diffusing out of the kernel. The pyrolytic carbon and silicon carbide layers provide an impenetrable barrier designed to contain the fuel and fission products. Some 15,000 of these coated particles, now about a millimeter in diameter, are then mixed with graphite powder and a phenolic resin and pressed into 50 mm diameter spheres. A 5 mm thick layer of pure carbon is then added to form a non-fuel zone, and the resulting spheres are sintered and annealed to make them hard and durable. Finally, the spherical fuel pebbles are machined to a uniform diameter of 60 mm (~2.4 inches). Each fuel pebble contains 9 g (~0.3 oz) of uranium. The total uranium in one reactor fuel load is 4.1 metric tons, and the total mass of a fuel pebble is 210 g (~0.5 pounds). During normal operation, the PBMR core contains a nominal load of 452,000 pebbles. A central graphite reflector column is located in the center of the core and an outer graphite reflector on the outside. The fuel core is in the annulus between the outer and central reflectors. Graphite is used in the reactor core because of its structural characteristics and its ability to slow down 3-47

76 neutrons to the speed required for the nuclear fission reaction to take place. The core and core structures geometry used in the PBMR provides inherent characteristics which limit the peak temperature in the fuel following an accidental loss of active cooling. The reactor is continuously replenished with fresh or reusable fuel from the top, while used fuel is removed from the bottom. After each pass through the reactor core, the fuel pebbles are measured to determine the amount of fissionable material left. If the pebble still contains a usable amount of the fissile material, it is returned to the reactor at the top for a further cycle. Each pass through the core takes about six months. Each pebble passes through the reactor about six times and lasts about three years before it is spent, which means that a reactor will use 15 total fuel loads in its design lifetime of 40 years. The extent to which the enriched uranium is used to the point where it is no longer of use in the core (called the burn-up level) is much greater in the PBMR than in the present nuclear power reactors. The remaining fissile material that could be extracted from spent PBMR fuel is of no use for proliferation purposes. This, coupled with the level of technology of coated particle fuel reprocessing, protects the PBMR fuel against the possibility of nuclear proliferation or other covert misuse. The fuel is transported to the spent fuel storage facility in the reactor building by means of a pneumatic fuel handling system. The spent fuel storage consists of 10 tanks, each with a diameter of 3.2 m (~10.5 ft.) and a height of 18 m (~59 ft.). One tank can store 600,000 pebbles (Also refer to Figure 3-13). The South African Nuclear Energy Corporation (Necsa) at Pelindaba near Pretoria, where fuel elements for Eskom s Koeberg nuclear reactor near Cape Town were manufactured in the past, is currently under contract from the PBMR project team to develop the fuel manufacturing capability using the technology established in Germany. All the main processes to manufacture kernels, coated particles and fuel spheres on laboratory scale are in place in the PBMR fuel development laboratories at Necsa. An Advance Coater Facility has been designed and is under construction. The first uranium-containing (depleted uranium) fuel spheres have been manufactured. The fuel development laboratories include quality control laboratories to test the product produced in the process development laboratories. The plan is to manufacture, during the first half of 2007, a small number of fuel spheres for irradiation testing that confirm PBMR capability to manufacture fuel spheres equivalent to the German reference fuel. In March 2005, PBMR (Pty) Ltd awarded a contract for the design, procurement, construction and cold commissioning of the pilot fuel plant utilities and infrastructure to Uhde, a South African division of Germany s Thyssenkrupp Engineering (Pty) Ltd. The facility will have an initial capacity of 270,000 fuel spheres per year. 3-48

77 PBMR Safety Features In all existing power reactors, safety objectives are achieved by means of engineered, active safety systems. In contrast, the PBMR is inherently safe as a result of the design, the materials used, the fuel, and the physics involved. This means that should a worst-case accident scenario occur, no human intervention would be required in the short or medium term. Nuclear accidents are principally driven by the residual power generated by the fuel after the chain reaction is stopped. This residual power (decay heat) is caused by radioactive decay of fission products. If this decay heat is not removed, it will heat up the nuclear fuel until its fission product retention capability is degraded and its radioactivity is released. In conventional reactors, the heat removal is achieved by active cooling systems (such as pumps), which rely on the presence of the heat transfer fluid (e.g. water). Because of the potential for failure in these systems, they are duplicated to provide redundancy. Other systems, such as a containment building, are provided to mitigate the consequences of failure and to act as a further barrier to radioactive release. In the PBMR, the removal of the decay heat is independent of the reactor coolant conditions. The combination of the very low power density of the core (one-twentieth of the power density of a Pressurized Water Reactor), and the resistance to high temperature of fuel in billions of independent particles, underpins the superior safety characteristics of this type of reactor. The helium, which is used to transfer heat from the core to the power-generating gas turbines, is chemically inert. It cannot combine with other chemicals and is non-combustible. The design objective is to ensure that the probability of air entering the primary circuit and corroding the high temperature core and graphite core structures is low. The peak temperature that can be reached in the core of the reactor is below the temperature that will cause damage to the fuel. This is because the radionuclides, which are the potentially harmful products of the nuclear reaction, are contained by two layers of pyrocarbon and a layer of silicon carbide that are extremely good at withstanding high temperatures. Furthermore, the fraction of the core that sees the peak temperatures is very small, further limiting any substantial fuel damage since individual fuel particles are independent of one another. Even if there is a failure of the active systems that are designed to shut down the nuclear reaction and remove core decay heat, the reactor itself will inherently shut down and eventually cool down naturally. Unlike the Chernobyl type of reactor, which during the accident produced more energy the hotter it became (known as a positive temperature coefficient of reactivity ), the pebble bed reactor has a strong negative temperature coefficient of reactivity, which stops the chain reaction. It also cools naturally by heat transport to the environment. The size and form of the PBMR core ensure a high surface area to volume ratio. This means that the high heat capacity of the core and core structures, together with the heat loss characteristics of the core (via the same process that allows a cup of tea to cool down) and the characteristics of the heat generated by the decay of fission products in the core, will limit the fuel temperature to below that value at which significant degradation of the activity retention capability can occur. The plant can never be hot enough for long enough to cause damage to the fuel. 3-49

78 Analysis shows that the progression of the fuel temperature as a function of time after a depressurized loss of forced cooling event took place is such that the maximum temperature will remain below that which will result in damage to the fuel. This inherently safe design of the PBMR renders obsolete the need for the typical safety backup systems and most aspects of the off-site emergency plans required for conventional nuclear reactors. It is also fundamental to the cost reduction achieved over other nuclear designs. The reactor core concept is based on the well-tried and proven German AVR power plant, which ran for 21 years. This safe design was proven during a public and filmed plant safety test, when the flow of coolant through the reactor core was stopped and the control rods were left withdrawn just as if the plant were in normal power generation mode. This test demonstrated that the nuclear reactor core shut itself down inherently within a few minutes. It was subsequently proven that there was no deterioration over and above the normal design failure fraction of the nuclear fuel. This proved that a reactor core meltdown was not credible, and that an inherently safe nuclear reactor design had been achieved. The reactor is housed in a building, part of which is a strengthened enclosure around the main power system. The module building comprises the entire structure that houses the power plant (excluding the generator), and is designed to withstand significant external forces such as aircraft impacts, tornadoes or explosions caused by saboteurs. The thickness of the reinforced concrete roof and walls (above ground level) of this structure is 1 m (~3.28 ft.). Within the module building is the reinforced concrete containment or citadel that encloses the reactor pressure vessel and the power conversion unit (excluding the generator). The walls surrounding the reactor pressure vessel are 2.2 m (~7.2 ft.) thick. The power conversion unit comprises the high- and low-pressure compressor unit, the power turbine, gearbox, generator, the recuperator and coolers. [35, 36, 37, 38, 39, 40] Design Enhancement for the PBMR One significant disadvantage of using PBMR technology is that the fuel is encased in potentially flammable graphite. To prevent fires, the reactor vessel is purged of oxygen. In addition, all current designs include an outer layer of silicon carbide on each fuel pebble to form a fireproof containment for each fuel pebble. The PBMR has several advantages over conventional water cooled reactor designs. The primary advantage is that its pebble bed fuel design is inherently self-controlling. As the fuel heats a proportion of the neutrons released by fission will be in a higher than normal energy spectrum through the Doppler effect. This higher energy spectrum allows some of the neutrons to react with the 238 U that is present in the fuel; thus, decreasing the neutron reaction rate in the 235 U and thus lowering the temperature of the fuel again since 238 U fission releases a negligible amount of energy compared to the energy released by 235 U fissions. Even if there is a failure of the active systems that are designed to shut down the nuclear reaction and remove core decay heat, the reactor itself will inherently shut down and eventually cool down naturally. Unlike the 3-50

79 Chernobyl type of reactor, which during the accident produced more energy the hotter it became (known as a positive temperature coefficient of reactivity ), the pebble bed reactor has a strong negative temperature coefficient of reactivity, which stops the chain reaction. It also cools naturally by heat transport to the environment. This inherently safe design of the PBMR renders obsolete the need for the typical safety backup systems and most aspects of the off-site emergency plans required for conventional nuclear reactors. Another significant advantage with using a PBMR is that it operates at higher temperatures than a conventional light-water reactor and can directly heat fluids for low pressure gas turbine use. This allows a turbine to extract more mechanical energy from the same amount of thermal energy. The PBMR Main Power System utilizes a recuperative Brayton cycle with helium as the working fluid. Helium at a relatively low pressure and temperature is compressed by a Low- Pressure Compressor (LPC) to an intermediate pressure, after which it is cooled in an intercooler. The intercooling between the two multistage compressors improves the overall cycle efficiency. The High-Pressure Compressor (HPC) then compresses the helium. The helium is preheated in the recuperator before entering the reactor that heats the helium to 900 C (1,652 F). After the reactor, the hot high-pressure helium is expanded in the Power Turbine, directly driving the LPC and HPC. The excess power is used to drive the generator via the gearbox. The still hot helium is cooled in the recuperator, after which it is further cooled in the pre-cooler. This completes the cycle. The heat rejected is equal to the heat transferred to the helium. The recuperator uses heat from the cooling process that would otherwise be lost to the main heat sink to heat the gas before it enters the reactor, thereby reducing the heating demand on the reactor and increasing the overall plant efficiency. The complex steam management system of light-water reactors is eliminated while the transfer efficiency is increased. Other advantages of using the PBMR technology include less radioactive fluids, self fuel containment, simpler design, and fuel variety (thorium, plutonium, unenriched uranium, or enriched uranium may be used), passive cooling design, no mechanism for a radioactive material release to the environment and no need for a EAB (Exclusion Area Boundary). The following is a listing of the design enhancement for the PBMR over the past pebble bed reactor designs and light water reactor designs: 1. An air tight containment building is not needed. There is not a mechanism during an accident that would lead to a release of radioactivity. Thus, a loss of coolant event is not a problem since the safety of the PBMR is not dependent on the presence of the helium coolant. (Note, the U.S. NRC has not approved a reactor design without a containment building). 2. The PBMR uses passive heat transfer mechanisms that do not require the helium coolant. The plant cools by natural convective, conduction and radiation (passive design). Also the core geometry provides for short heat transfer paths to outside of the reactor pressure vessel. Thus, the reactor thermal power and core configuration are designed to assure passive decay heat removal without fuel damage during hypothetical accidents. 3. Since the control rods are in the reflector around the core of fuel balls (pebbles) the controls rods can not damage the fuel pebbles as they are moved. [40] 3-51

80 The mission of the NRC is to license and regulate the Nation's civilian use of byproduct, source, and special nuclear materials to ensure adequate protection of public health and safety, promote the common defense and security, and protect the environment. Thus, one of its primary purposes and responsibilities of the NRC is to determine whether a reactor design is safe and can be built in the United States. The design certification process is currently being used for the new reactor designs in the United States. The PBMR design currently does not have NCR certification Estimated Cost PBMR (Pty) Ltd. is currently reviewing its cost projections due to the considerable increases in global materials and equipment prices. Revised estimates will be available in early However, the following costs are provided, based on published literature, as an estimate of potential PBMR costs. Comparing overnight capital cost of small modular plants with short construction durations, smaller number of operating staff and less demanding site conditions with large LWR s is inappropriate. However, overnight construction cost (when in clusters of eight units) is expected to be $1,000/kWe and generating cost below 3 cents/kwh. It is intended to build these units in groups of four (4) or eight (8) modules. Follow on modules would be built one after the other in approximately six (6) month periods after the lead module is completed. Assuming a $1,000/kWe overnight cost for a 175 MWe unit would mean the unit would have an overnight construction cost of $175 million. These costs could grow to $1,300/kWe once the owners cost and the AFUDC are included ($230 million), see Table 3-10 for cost estimates (Note: These cost estimates were not directly provided by the reactor vendor and are not endorsed by the reactor vendor. The cost estimates were developed based on the information that was publicly available. No guarantee is provided as to the accuracy of these values. For example, these values would change as a result of increasing material, equipment and labor prices). [9, 13, 35] 3-52

81 Table 3-10 Technology Basis Data (PBMR) Plant Type PBMR Technology Provider by PBMR (Pty) Ltd. (South Africa) Enriched Uranium Dioxide, currently without reprocessing. Fuel Type (Can utilize combination of thorium and plutonium fuel) Ceramic spheres covered in graphite [38] Currently there are no companies manufacturing PBMR fuel in the U.S. Fuel Cycle Number of Units Unit Size, MW On line refueling 1 [38] Note: it is intended to build these units in groups of eight (8) modules. 400 MWt [38] 175 MWe (gross) [40] Plant Size, Net MWe 160 to 165 Output to the grid [41] Plant Auxiliaries, MWe 10 to 15 (175 MWe (gross) 160 to 165 MWe (net) = 15 to 10 MWe) Plant Capacity Net 94% Available for Commercial Orders, Year First Commercial Service (Technology Year) Hypothetical In-Service Year Design & Cost Estimate Rating 2012 [35] 2015 [35] 2012 Demonstration Plant 2015 First complete Eskom 4 module plant [35] E - Goal Technology Development Rating D Pilot (Demonstration plant planned in South Africa) [38] Plant Location The Koeberg facility near Cape Town, South Africa [38] Plant Capital Cost 1 Unit Structures & Improvements Reactor Plant Equipment Turbine Plant Equipment Electrical Plant Equipment $20 million ($125/kWe) $40 million ($250/kWe) $20 million ($125/kWe) $10 million ($63/kWe) 3-53

82 Table 3-10 Technology Basis Data (PBMR) (continued) Plant Type PBMR Technology Provider by PBMR (Pty) Ltd. (South Africa) $3 million ($19/kWe) for miscellaneous plant equipment Other BOP Facilities $2 million ($13/kWe) for the Main Condenser Heat Rejection System $10 million ($63/kWe) for construction services, engineering / home office services and field supervision / field office services Installation General Facilities & Engineering Fee Project & Process Contingency Total Plant Cost Total Cash Expended (Mixed year $) AFUDC (interest during construction) Total Plant Investment (includes AFUDC) Total Owner Costs Total Capital Requirement, Hypothetical In-Service Year (includes AFUDC) $40 million ($250/kWe) manual labor $30 million ($188/kWe) Non manual labor See Other BOP Facilities above $20 million ($/125kWe) (12% of the plant capital cost of $175 million) $195 million ($1,219/kWe) To be supplied by EPRI $15 million ($94/kWe) (Based on 8% interest, 8 million per month for 24 months) $210 million ($1,313/kWe) $20 million ($125/kWe) (Based on 10% of the total plant cost) To be supplied by EPRI Cost for Hypothetical In-Service Year Operating & Maintenance Costs Fixed $/kw-yr Information not available Incremental, mils/kwh: Variable (includes consumables), mills/kwh 4.64 fuel cost 2.45 variable 7.09 total [10] 3-54

83 Table 3-10 Technology Basis Data (PBMR) (continued) Plant Type PBMR Technology Provider by PBMR (Pty) Ltd. (South Africa) Net Plant Heat Rate, Btu/kWh (if applicable) Full Load 8,280 [35] Min Load 8,570 (40% [10]) Average Annual Information not available Unit Availability Equivalent Planned Outage Rate, % Equivalent Unplanned Outage Rate, % ~2 General maintenance period: 30 to 50 days scheduled per 6 years [35] Equivalent Unplanned Outage Rate, <1 Total availability target of > 95 [35] Equivalent Availability, % > 95 Capability Ratio Total availability target of > 95% [35] Duty Cycle Minimum Load, % 40% [35] Preconstruction, License & Design Time, Years Idealized Plant Construction Time, Years 4 to 6 years (for the U.S.) 2 years for the N th commercial plant first module of a multi-module plant with as little as 6 months for follow-on modules. Actual schedule depends on owner requirements to match power with demand. [35] Unit Life, Years [35, 40] 125 (minimum) Estimated Land Area Needed (acres) Estimated Water Requirements Excluding cooling ponds The EAB would be about 0.25 miles from the reactor building. A module itself (power block) would only occupy approximately 1.4 acres 4,000 gpm closed loop Once through cooling: 6 kg/s per MWt (~38,000 gpm for 400 MWt) 1. Because no new nuclear plant has been constructed in the U.S. in the last two decades and due to the recent ( ) cost increases, there is quite a bit of uncertainty in the estimated costs. 2. While the estimation of plant costs is an art rather than a science and the site-specific costs are unique to each site, the estimates provided here are based on sound design basis and cost estimating basis to serve as a baseline. 3-55

84 3.4.4 Deployment Plans Summary In 2003, the South African government approved construction, subject to local regulatory processes, of a PBMR demonstration plant at the Koeberg facility near Cape Town, South African and a pilot fuel plant near Pretoria, South Africa. The PBMR plant will be built by PBMR (Pty) Ltd. The current major investors in PBMR (Pty) Ltd include the South Africa Government, Eskom, the Industrial Development Corporation (IDC) of South Africa, and Westinghouse. In 2005, PBMR (Pty) Ltd awarded a contract for engineering, procurement, and construction management to SLMR (a Canadian-South Africa joint venture) for the demonstration plant. Construction for the demonstration plant is planned to start in 2007 and for the plant to be complete in 2011, while the first commercial plants are planned for If the trial of the PBMR demonstration plant at the Koeberg facility proves successful, PBMR (Pty) Ltd will plan to build up to 30 plants in South Africa. PBMR (Pty) Ltd also has a future plan of exporting up to 20 plants per year, providing a very useful addition to the South African economic market. PBMR Process Heat Plant designs are being developed for deployment after 2018 as a second product line. The fundamental safety and reactor design are derived from the basic electric plant design developed from the Demonstration Power Plant and adapted to an indirect cycle through intermediate heat exchangers. Various test facilities are being constructed and commissioned as part of the technology development in South Africa. These facilities include the Pebble Bed Micro Model (PBMM), Helium Test Facility (HTF) and the Heat Transfer Test Facility (HTTF). The main objectives of these facilities are: Research and Development Non-Nuclear Code Verification and pre-qualification Technology Demonstration Skills development The PBMM is considered to be the first closed-cycle, multi-shaft gas turbine in the world and was completed in The main objectives of the project were to prove that a multi shaft recuperated Brayton cycle can be sustained and controlled, that it renders a stable operating configuration and to provide code verification. The four control philosophies of the PBMR identified for demonstration were start-up, full power operation, load following and load rejection. The HTF team is busy with commissioning. This facility will be used to test the following systems and sub-systems of the Demonstration Power Plant: Fuel Handling and Storage System (FHS), Reactivity Control and Shutdown Systems (RCS and RSS), Helium Inventory Control and Blow Down System (HICBS), 3-56

85 Gas Cycle Valves (GCV), Helium Purification Facility (HPF), Instrumentation, and Main Power System (MPS) Recuperator The HTTF has two test units, namely the High Pressure Test Unit (HPTU) and the High Temperature Test Unit (HTTU). As a whole the purpose of HTTF is to perform separate effects and integrated effects tests to determine heat transfer characteristics within an annular pebble bed. The HPTU focused on providing separate effects test data and HTTU focused on integrated effects tests. The HPTU was completed in August 2006 and PBMR s test program is currently being executed on this facility. The HTTU completion is planned for middle of [35, 36, 37] 3-57

86 Table 3-11 Technology Monitoring Guide (PBMR) Leading Developers of the Science or Technology Unresolved Issues Changes To Watch For Major Trends Leading Vendors Industrial Firms Nonprofit Organizations Government Organizations R&D Intensity Technologies PBMR On going South African Government[38] none Industrial Development Corporation (IDC) of South Africa [38] PBMR (Pty) Ltd. [41] Westinghouse [38] Mitsubishi Heavy Industries Design certification by the U.S. NRC. Process Heat Applications and Desalination RSA Licensing 3-58

87 Table 3-12 Technology Development and Assessment (PBMR) Features and Characteristics of Technology Major Technical Issues Key Vendor Activities Resource Requirements That Impact Technology Key Business and Market Indicators Current Status High safety, small size, great environmental hygiene, short construction periods, costcompetitiveness [38] Establishing modern licensing rules for modular HTGR designs using riskinformed, performance-based criteria. Establishing reactor graphite design and inspection codes under ASME. Direct power conversion cycle Proliferation resistance fuel design. [35] Building a Demonstration Power Plant [35] None [35] Confirmation of equipment reliability and analytical code verification and validation. PBMR (Pty) Ltd has several large integrated test facilities including the Helium Test Facility for long term equipment reliability; the High Temperature Test Facility for code V&V; the PBMR Micro Model for Brayton Cycle performance and control verification; plus numerous component prototypes and test platforms. Additionally, PBMR (Pty) Ltd is manufacturing early reactor fuel samples for irradiation testing that demonstrate the effective transfer of the original German fuel manufacturing processes to South Africa. [35] Future Considerations or Trends Information not available Information not available Completion and startup of Eskom 4 Information not available Information not available 3-59

88 Table 3-12 Technology Development and Assessment (PBMR) (continued) Current Status Future Considerations or Trends Key technology needs Information not available Information not available Technology Outlook Development Timeframe Eskom South Africa s state utility has given a letter of intent for 4 GW of PBMR capacity. [35] Loading fuel for the DPP is planned for [35] Information not available Information not available Research Ongoing Information not available Development Ongoing Information not available Demonstration Projected Commercialization Date South Africa Demonstration Plant - Start Construction 2007, Completed by [38] As part of commercialization the first 4 unit plant is planned after completion of the DPP. Construction is expected to begin in Fuel load for the first unit is expected in Loading fuel for the fourth unit is planned for [35] Information not available Information not available The Demonstration Power Plant is currently being licensed through the National Nuclear Regulator in South Africa. The design is being established with international codes and standards and regulatory requirements for ease of deployment outside of South Africa. PBMR (Pty) Ltd notified the NRC (2004) that it intended to apply for design certification in the future and requested discussions with the NRC to plan the scope and content of the pre-application review. NRC staff has held several public meetings with PBMR (Pty) Ltd to discuss its activities and plans to submit pre-application information. PBMR (Pty) Ltd has continued to submit pre-application information through 2006 and expects to submit a design certification application in late Design Certification completion is expected in following the Demonstration Power Plant operation. [35, 44] 3-60

89 3.4.5 Installations Under Construction or Planned In 2003, the South African government approved construction of a PBMR demonstration plant at the Koeberg facility near Cape Town, South Africa and a pilot fuel plant near Pretoria, South Africa (See Section ). In 2004 an important milestone for PBMR (Pty) Ltd was reached with the construction of a helium test facility (HTF) at Pelindaba, South Africa. This facility is being built for the testing of the complete helium cycle system for the PBMR. The HTF is also designed to simulate the reactivity control, fuel-handling, and shutdown systems of the PBMR. Currently, there are no potential combined license applications for the PBMR scheduled with the U.S. NRC. [20, 30, 37, 38] U.S. Design Differences For the DPP the turbine shaft rotates at 100 Hz and drives the generator through a gearbox at 50 Hz. For the U.S. a different gearbox will be used to supply power at 60Hz. This avoids redesigning the power turbine and turbo-compressor units. BOP power changes for different voltage and frequency of electrical equipment is not required beyond the first U.S. unit. The properties of low enriched uranium TRISO coated pebbles are very important factors for determining the radiological safety of the PBMR. This is because the fission product retention in the fuel pebbles and the maximum fuel temperature that can be tolerated in the reactor core, are determined by the coated pebble properties. Historically, it has been found that the in-reactor performance of U.S. made tristructuralisotropic (TRISO) coated pebbles is significantly lower than that of German-made TRISOcoated pebbles for gas reactor particle fuel. In particular, it can generally be found that German fuel shows in-reactor gas release values 1,000 times lower than that of U.S. fuel. The German fuel is also found to be of a higher quality of production, and to have higher safety test behaviors. The German fuel comparison comes from the program supporting the German AVR (Arbeitsgemeinschaft Versuchsreaktor-working group test reactor) and THTR (thorium hightemperature nuclear reactor) reactors, while that of the U.S. is from the U.S. program post Ft. St. Vrain. Several design differences are attributed to causing this difference of in-reactor quality between U.S. and German designed TRISO-coated pebbles. One such design difference is a technical difference in the microstructures of the PyC and SiC layers in the TRISO coating and the bonding of those layers. This difference in microstructures also relates to differences in the fabrication processes used in the U.S. and German design. Another design difference comes from the production goals behind the U.S. and German facilities. The German design had a focus on UO 2 -TRISO fuel form since fabrication was focused on an industrial/production scale. In contrast, U.S. fabrication was for a mixture of lab and large scale, supplying many different variants of TRISO-coated fuel. [35, 42, 43, 44, 45, 46, 47, 48, 49] 3-61

90 3.5 AREVA NP, an AREVA and Siemens Company Large Advanced Evolutionary Nuclear Reactor (U.S. EPR Not the Finnish or French Design) The U.S. Evolutionary Power Reactor (U.S. EPR) is a 1,600 MWe (nominal) Generation III+ evolutionary power reactor based on proven PWR technology. The U.S. EPR is being designed for the U.S. by AREVA Inc. and is based on AREVA s advanced nuclear power plant (i.e. its own EPR) Plant Description The U.S. EPR has 241 fuel assemblies surrounded by a neutron reflector to optimize fuel utilization and protect the pressure vessel from radiation damage. The plant has a lifetime of 60 years and an average capacity factor of 94%. The key features of this advanced PWR are an improvement in economy, safety, and reliability. The plant is designed to cost 10% less to operate than most of the conventional plants in service today. The U.S. EPR has been greatly simplified as compared with existing plants. The plant has 47% fewer valves, 16% fewer pumps, 50% fewer tanks, and 44% fewer heat exchangers than the current PWR design. The plant design also has only those features and materials that have shown superior performance over the last 40 years of nuclear power plant operation, improving both reliability and operation and maintenance costs. The reactor can use various types of fuel, such as low enriched uranium (i.e. 5%) or MOX (mixed-oxide) fuel. The EPR design also allows for a flexible operating cycle (i.e. 12 to 24 months). Another feature of the U.S. EPR that makes it very reliable is that many maintenance and inspection tasks can be completed while the reactor is operating. This in turn also minimizes downtime and maximizes plant efficiency (i.e %). The EPR has four pressurized water coolant loops. The Reactor Coolant System (RCS) is composed of the reactor vessel that contains the fuel assemblies, a pressurizer including control systems to maintain system pressure, one Reactor Coolant Pump (RCP) per loop, one steam generator per loop, associated piping, and related control and protection systems. The RCS is contained within a concrete containment building. The containment building is enclosed by a Shield Building with an annular space between the two buildings. The pre-stressed concrete shell of the Containment Building is furnished with a steel liner and the Shield Building wall is reinforced concrete. Figure 3-15 shows the plant layout of the U.S. EPR and Figure 3-16 shows a simplified flow diagram of the EPR. The Containment and Shield Buildings comprise the Reactor Building. The Reactor Building is surrounded by four Safeguard Buildings and a Fuel Building. The internal structures and components within the Reactor Building, Fuel Building, and two Safeguard Buildings (including the plant Control Room) are protected against aircraft hazard and external explosions. The other two Safeguard Buildings are not protected against aircraft hazard or external explosions; however, they are separated by the Reactor Building, which restricts damage from these external events to a single safety division. 3-62

91 Redundant 100% capacity safety systems (one per Safeguard Building) are strictly separated into four divisions. This divisional separation is provided for electrical and mechanical safety systems. With four divisions, one division can be out-of-service for maintenance and one division can fail to operate, while the remaining two divisions are available to perform the necessary safety functions even if one is ineffective due to the initiating event. In the event of a loss of off-site power, each safeguard division is powered by a separate Emergency Diesel Generator (EDG). In addition to the four safety-related diesels that power various safeguards, two independent diesel generators are available to power essential equipment during a postulated Station Blackout (SBO) event loss of off-site ac power with coincident failure of all four EDGs. Water storage for safety injection is provided by the In-containment Refueling Water Storage Tank (IRWST). Also inside containment, below the Reactor Pressure Vessel (RPV), is a dedicated spreading area for molten core material following a postulated worst-case severe accident. The safety approach for the protection of the core relies partly on the capacity to predict and measure the nuclear power level (or level of neutron flux) as well as the three-dimensional power distribution. The measurement of the nuclear power level (or neutron flux level) is performed by loop temperature instrumentation and wide range excore flux (power) instrumentation, as is classically done on PWRs. The capacity to predict and measure the three-dimensional power distribution relies on two types of in-core instrumentation: Movable instrumentation (aeroball fast flux mapping measurement system) Fixed instrumentation used for delivering the necessary on-line information to the different core surveillance and core protection systems The aeroball system utilizes stacks of vanadium alloy steel balls, inserted from the top of the reactor vessel, that are pneumatically transported into the reactor core inside guide thimbles of the fuel assemblies. This system is simple and reliable. The guide tubes for the balls have a small inner diameter ( in); their bend radii are small (only a few inches); and there are no major constraints for locating the measurement room and routing the tubing. The time periods necessary for a flux measurement are 3 minutes for activation followed by 5 minutes for activity measurements. This system therefore allows flux-mapping measurements in time intervals of 10 to 15 minutes. Figure 3-17 shows a schematic of the aeroball system for the U.S. EPR. This technique of flux-mapping, combined with the short half-life of the V 52 isotope (3.7 min.) that serves as the indicator, restricts the range of power over which accurate three-dimensional flux mapping is possible. In practice, acceptable two-dimensional flux maps can be obtained at approximately 5% reactor power and accuracy necessary for three-dimensional flux maps is reached at approximately 30% reactor power. With the fixed in-core measurements, periodically calibrated by the very quick movable aeroball system, the on-line core surveillance and protection is far more precise and contributes to a significant increase of the safety margins than from excore instrumentation. The aeroball system has operating experience in Siemens plants since

92 The fuel pool is located outside the Reactor Building in a dedicated building to simplify access for fuel handling during plant operation and handling of fuel casks. The Fuel Building is protected against aircraft hazard and external explosions. Fuel pool cooling is assured by two redundant, safety-related cooling trains. Each train consists of two pumps installed in parallel, a heat exchanger cooled by the CCWS (Component Cooling Water System), and associated piping and valves. The pipe penetrations to the spent fuel pool are above the required level of water that must be maintained over the spent fuel, while providing the required pump suction head. The pipes that penetrate the pool are equipped with siphon breakers to limit water loss resulting from a leak in the piping system. [50, 51, 52, 53, 54, 55] Figure 3-15 Sample Plant Layout of the U.S. EPR (Provided by AREVA NP) 3-64

93 Figure 3-16 Simplified Process Flow Diagram for the EPR (Provided by AREVA NP) [50] 3-65

94 Figure 3-17 Aeroball System Schematic for the U.S. EPR (Provided by AREVA NP) [55] 3-66

95 Figure 3-18 Example of the Aeroball System Configuration in the Reactor [52] 3-67