3 project descriptionprojects.gibb.co.za/portals/3/projects/200911 pmbr/j27196... · 2017. 11....

ESKOM PBMR DPP EIA Version 1.0 / October 2009 Environmental Impact Report

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3 PROJECT DESCRIPTION

3.1 Introduction

Eskom is seeking environmental authorisation for the construction, commissioning, and operation of a 400 MW(t)20 Pebble Bed Modular Reactor Demonstration Power Plant (PBMR DPP). The proposed location for the PBMR DPP is approximately 400 m to the south of the existing Koeberg Nuclear Power Station on the Farm Duynefontein 34. The footprint of the PBMR DPP in the operational phase of the project, inclusive of all associated infrastructure, will be approximately 9 hectares (ha). In the event that the proposed project is authorised, it is estimated that the construction of the PBMR DPP could commence in 2010 with commissioning of the plant in 2016. The following section provides a description of the project including the location, infrastructural requirements, transportation considerations and technical specifications of the PBMR DPP. The safety, licensing and decommissioning of the plant will be discussed and an overview of nuclear energy, the nuclear fuel cycle and the history and function of nuclear technology and the PBMR in particular, will also be provided in this section.

3.2 Location of the Proposed Site for the PBMR DPP

The Koeberg Nuclear Power Station (KNPS) site is located within the Eskom Controlled Area on the farm Duynefontein (farm number 34), within the Koeberg Private Nature Reserve, the latter being approximately 3000 ha in extent. The KNPS site is approximately two (2) km from the Duynefontein residential area, 30 km north of Cape Town and 10 km south of Atlantis, within the City of Cape Town Metropolitan Municipality jurisdiction (Figure 3-1) . The PBMR DPP is proposed to be located some 400 m south of the existing Koeberg Power Station, inside the Access Control Point (ACP) 1 security fence of the KNPS site (Figure 3-2) . The PBMR DPP would require approximately 9 ha of the KNPS site, which is approximately 125 ha in extent. The site and surrounding nature reserve are managed according to a formal integrated environmental management system.

20 The unit of power in the form of heat produced in the reactor core.


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Figure 3-1: Regional location of the proposed site for the PBMR DPP


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Figure 3-2: Proposed location of the PBMR DPP on th e KNPS site


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The area within the existing KNPS site boundary forms the main study area for the PBMR DPP. This study area was assessed by all specialists, however, some specialist studies were required to extend beyond this area for the purposes of their assessment. The table below lists the specialist studies and the extent of the study area that was evaluated by each study for the purpose of this EIA.

Table 3-1: Specialist Studies and Extent of Study A rea Evaluated by the Specialists

Specialists Studies Extent of study area evaluated by specialist. Meteorology Study The Mediterranean Climate Region of the Western Cape Province and

microclimatology of the Koeberg Area Geology and Seismic Risk The area within a 320 km radius of the KNPS Surface and Ground Water Study

Study area of 10 km radius around the Site

Wetland Study The study area has been assessed at a number of levels, comprising: • the proposed PBMR DPP site itself and associated construction / lay

down areas including the site east of the R27; • the Koeberg Nuclear Site as a whole, including the Koeberg Nature

Reserve; • other freshwater systems in the broader area, including the lower Sout

River; and • the Modder River to the north – potentially affected by the proposed

transportation of PBMR DPP equipment during the construction phase. Epidemiology Study 18 km radius around the Koeberg site Air Quality Study The dispersion of pollutants was modelled for an area covering 30 km

(north-south) by 30 km (east-west). The proposed PBMR DPP site was placed in the centre of the modelling domain.

Fauna Study • The proposed PBMR DPP site and adjacent laydown area south of the present power station;

• Alternative laydown areas to the north and north-east of the Koeberg Nuclear Power Station, to the north of the conservation office and to the south of the conservation office;

• The Triangle to the east of the R27; • Access roads and powerlines; • The Water pipeline between the R27 and the proposed PBMR site;

and • The Modder River crossing.

Flora Study • The proposed PBMR DPP site and adjacent laydown area south of the present power station;

• Alternative laydown areas to the north and north-east of the Koeberg Nuclear Power Station, to the north of the conservation office and to the south of the conservation office;

• The Triangle to the east of the R27; • Access roads and powerlines; • The Water pipeline between the R27 and the proposed PBMR site;

and • The Modder River crossing.

Marine Fauna and Flora Study

The southern limit of the relatively uniform Namaqua marine biogeographic region, which extends north as far as Luderitz

Site Safety and Security Study

The existing KNPS site

Emergency Response Report

The area within an 80 km radius of the existing KNPS

Socio-Economic Study The City of Cape Town, the southern most metropolitan area in South Africa, serves as the focus of this study.

Land Use Study The study area includes land within a 10 km radius from the proposed PBMR DPP site, covering an area of approximately 17 691 hectares.

Traffic Study Transport networks within the existing KPNS site and the surrounding areas, including within the 16 km Urgent Protective Action Zone (UPZ), and from Saldanha Bay Harbour, Vaalputs and Pelindaba to the KPNS site.

Heritage Study • The proposed PBMR DPP site and adjacent laydown area south of the present power station;


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Specialists Studies Extent of study area evaluated by specialist. • Alternative laydown areas to the north and north-east of the Koeberg

Nuclear Power Station, to the north of the conservation office and to the south of the conservation office;

• The Triangle to the east of the R27; • Access roads and powerlines; • The Water pipeline between R27 and proposed PBMR site; and • The Modder River crossing.

Noise Study The study area encompasses the residential suburb of Duynefontein, with the nearest residences approximately 1800 m south of the existing nuclear plant and approximately 1200 m south of the proposed PBMR building. This is the closest occupied noise-sensitive land that might be affected by the proposed PBMR.

Visual Impact Study This report considers the visibility or views of the project within a radius of approximately 15 km from the PBMR DPP and Construction Laydown Area.

Oceanography Study The marine environment immediately adjacent to the KPNS site. Helium Study The study did not have a physical boundary, as it addresses issues

pertaining to the cost and availability of helium gas. Radiological Health and Safety Report

The study did not have a physical boundary. Rather, radiation emissions from existing operations and potential pathways of exposure are used to determine the possible area of influence and associated radiological effects on the receiving environment.

Financial Aspects This extent of this study is based on financial parameters rather than physical boundaries, and as such, is national in scope since it focuses on the financial provisions required for the construction, long term custody, decommissioning, dismantling and rehabilitation of the proposed PBMR DPP.

Radiological Waste Management Report

The extent of this study encompasses the existing KPNS site and Vaalputs. However, the scope of this study also addresses the manner in which nuclear fuel is currently transported to the KNPS and the manner in which nuclear fuel is likely to be transported to the proposed PBMR DPP, the international basis for the management of high-level waste and an overview of the radioactive waste management practices envisaged to being part of the radioactive waste management programme for the PBMR DPP, from generation to disposal.

Nuclear Energy Report National extent, as this study investigates the PBMR’s compatibility with the existing electricity generating system of South Africa and whether it would make any renewable energy electricity generating technology that exists now or may exist in the future, incompatible with South Africa’s electricity generating infrastructure.

3.3 Nuclear Energy

Nuclear energy can be regarded as a source of heat energy which is derived from the splitting of the nucleus of a uranium atom. The properties of different chemical elements are determined by how many protons there are in the nucleus of the atom (Appendix D) . There are just over 100 known different atoms, known as elements. Atomic nuclei become more unstable the larger they are, and they tend to break apart. This is known as “radioactive decay”. Uranium is the heaviest atom found in nature, its Atomic Number is 92. Uranium also undergoes radioactive decay, but very slowly, so much so that there are still large quantities of uranium to be found in the ground. Due to the fact that the uranium nucleus is so large it can be induced to undergo rapid splitting, known as “fission” in a nuclear reactor. It is this fission property that makes uranium such a good nuclear fuel. When the uranium atom splits it releases nuclear energy (Appendix D) .


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In the case of the PBMR DPP, the heat produced by the fission reaction is transferred to a helium coolant which is then used to drive a turbine that generates electricity. Other Nuclear Power Stations (NPSs) utilise the heat energy to produce steam from water. The steam is then used to drive a turbine to generate electricity.

3.4 History of Nuclear Power

3.4.1 The Period up to the 1960s 21

The first proven nuclear fission was performed in Germany by Otto Hahn in 1939. The political situation in the world during the Second World War demanded the rapid development of nuclear weapons technology in the United States. On 2 December 1942 researchers under the leadership of Enrico Fermi managed to maintain the first controlled nuclear chain reaction in an experimental device called Chicago Pile 1, which was a large composite of uranium and graphite. The reaction was maintained for 28 minutes and represents the basis of today's dominating technology of thermal nuclear reactors. In December 1951 the first four electric bulbs were illuminated by electricity produced by the nuclear reactor EBR-1 in Idaho Falls in the United States. This was the first experimental breeder reactor. There was a parallel development of the pressurized water reactor (PWR), which was first designed for the propulsion of submarines and successfully tested in 1953. In 1954 in Obininsk, in the Soviet Union, the first Nuclear Power Station (NPS) in the world was successfully made operational: APS-1. It produced only 5 MW of electric power and was graphite moderated and marks the beginning of the later Reaktor Bolshoy Moshchnosti Kanakniy (RBMK), which is translated as “Reactor of high power (of the) channel (type)”. On 17 July 1955 the experimental boiling reactor BORAX-III produced, for the first time, enough electricity to illuminate the city of Arco in Idaho, United States. On 17 October 1956 the first gas-cooled NPS was started in Calder Hall in Great Britain, which marked the beginning of the British nuclear programme and the first commercial scale use of electricity from nuclear power. The Calder-Hall plant was finally shut down in 2003 after 46 years of operation. The first commercial NPS in the USA, Shipping Port, was started in 1956. The second commercial NPS, Yankee-Rowe was built in 1960 in Massachusetts. Both NPSs used PWR technology and were constructed by Westinghouse. In 1960 the General Electric Company started the first commercial boiling water NPS, Dresden near Chicago in the United States of America. In 1962 the first Canadian NPS, NPD was commissioned, which used natural uranium and heavy water.

3.4.2 The Period of the 1960s – 1980s 22

Through the 1960s and 1970s nuclear power was a favoured resource for utilities needing new capacity. The size of units steadily increased with later units having typical capacities in the 1000 to 1200 MW range. By 1973 the US had built 22 700

21 Arcus GIBB (2008) 22 Appendix D


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MW of nuclear capacity. This capacity was expanded to 51 800 MW by 1981 and to 99 600 MW by 1990. However, in the US, nuclear utilities did not work together very well which meant that virtually every new nuclear station built was built from scratch as a unique unit, there was little coordination and standardisation to reduce cost. This resulted in many construction delays and construction costs being exceeded to the detriment of consumers. This compounded a negative public perception of nuclear energy. To make matters worse, the regulatory system was unwieldy, for example requiring a licence to build a nuclear plant and then a separate licence to operate it. This led to additional confusion and subsequent cost increases over the original projected project costs (Alexander L, 2004). But in the 1990s, the situation improved. Although in the US no new construction had taken place, the existing plants were being run in a much more cost-effective manner, finally bringing in the financial returns that had been promised when they had been built, i.e. Although construction costs had been high initially, over the long term, the lower operating and fuel costs would pay off.

3.4.3 The Period of the 1990s to Date 23 Today, the world produces as much electricity from nuclear energy as it did from all sources combined in 1960. Civil nuclear power can now boast over 12,000 reactor years of experience and supplies 16% of global needs. Currently about 30 countries have some 440 commercial nuclear power reactors with a total installed capacity of over 360 000 MW(e). This is more than three times the total generating capacity of France or Germany from all sources. Some 30 further power reactors are under construction, equivalent to 6% of existing capacity, while about 35 more are firmly planned, equivalent to 10% of present capacity (WNA, 2007a). Although fewer nuclear power plants are being built now than during the 1970s and 1980s, those now operating are producing more electricity. In 2004, production was 2 618 billion kWh. The increase over the last five years (218 Terawatt hours) is equal to the output from 30 large new nuclear plants. Yet between 1999 and 2004 there was a net increase of only two reactors. The rest of the improvement is due to better performance from existing units. In a longer term perspective, from 1990 to 2004, world nuclear energy capacity rose by 39 Gigawatts (12%), due both to net addition of new plants and upgrading some established ones and electricity production from nuclear energy rose by 718 billion kWh (38%). The relative contributions to this increase were: new construction 36%, uprating 7% and availability increase 57%. One quarter of the world's nuclear reactors have load factors24 of more than 90%, and almost two thirds do better than 75%, compared with about a quarter of them in 1990 (WNA, 2007 b). For 15 years Finnish plants topped the performance tables, but Japan, US and South Korean plants now dominate the top 25 positions.

23 Appendix D 24 The load factor of an electricity generator is the amount of electricity produced compared to its theoretical maximum

generation capacity.


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US nuclear power plant performance has shown a steady improvement over the past 10 to 15 years, and the average load factor now stands at around 90%, up from 65% in 1990. This places the US among the performance leaders with 12 US reactors in the top 25 reactors all achieving more than 98% load factor. The US accounts for nearly one third of the world's nuclear electricity. (WNA, 2007 b) Some of these load factor figures suggest near-maximum utilisation, given that most nuclear reactors have to shut down every 18 to 24 months for fuel change and routine maintenance.

3.4.4 South Africa’s first Nuclear Power Station 25 South Africa began with the construction of its first NPS known as the Koeberg Nuclear Power Station (KNPS) (owned and operated by the country's only national electricity supplier, Eskom) in 1976, with Unit 1 being synchronised to the grid on 4 April 1984. Unit 2 followed shortly thereafter on 25 July 1985. Koeberg is the only NPS in South Africa. It is located approximately 30 km north of Cape Town, near Melkbosstrand on the west coast of South Africa. It was anticipated that the plant would serve as the only base load power station in the Western Cape. The high costs associated with the transportation of fossil fuels from other areas of the country to this point rendered coal fired power stations in the Western Cape economically unfeasible. Koeberg has two uranium fuelled PWRs supplied by Framatome of France (under license from Westinghouse USA). The Koeberg NPS has two large turbine generators of over 900 MW with a total capacity to supply 1800 MW to the national grid (after internal consumption). The grounds of the nuclear plant form a 22 km² nature reserve open to the public containing more than 150 species of birds and half a dozen small mammal species. When Koeberg was commissioned, the Western Cape’s demand was less than that of Koeberg’s capacity. Thus the excess power generated by Koeberg was transmitted to other parts of South Africa. Subsequently, the Western Cape’s demand exceeded Koeberg’s capacity and currently approximately 3 000 MW is imported into the Cape region to accommodate the growth in demand.

3.4.5 Commercial Nuclear Power Technologies NPS differ from each other mostly by the type of nuclear technology used. The most common types of nuclear technology are outlined below26:

Pressurized Water Reactor (PWR) , which is moderated and cooled with light water27 that is not boiled in the reactor. The turbine is driven by steam from the steam generator;

Boiling Water Reactor (BWR) , which is moderated and cooled with light water. The water is boiled in the reactor, which generates steam that drives the turbine;

Gas Cooled Reactor/ Advanced Gas Cooled Reactor (GC R/AGCR), is moderated by graphite and cooled by carbon dioxide. Water is heated

25 Arcus GIBB (2008) 26 Arcus GIBB (2008) 27 Oridinary water composed of two hydrogen atoms and one oxygen atom.


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and boiled by hot carbon dioxide, this produces steam, which in turn drives the turbines;

Light Water Cooled Graphite Moderated Reactor (LWCG MR), is moderated by graphite and cooled with light water, which boils in the reactor to produce steam, driving the turbines; and

Pressurized Heavy Water Moderated Reactor (PHWMR) is moderated and cooled with heavy water28. Light water is heated in the steam generator to drive the turbines.

3.5 Development History of the PBMR Technology

The PBMR DPP is based on the designs developed as a result of an extensive high temperature reactor (HTR) development programme in Germany. Extensive research and development has been done on the 15 MW(e)29 [40 MW(t)] Arbeitsyemeinschaft Versuchs Reaktor (AVR) research reactor at the nuclear research centre in Jülich. The reactor operated from 1966 to 1988, when it was decommissioned due to political considerations, and because it had fulfilled all planned research tests and experiments. Although it was a prototype in test mode, it produced power for 70% of its life. During its 22 years of operation, the design proved the superior behaviour of the coated particle fuel concept, the favourable safety characteristics of the core, and even fulfilled the safety requirements listed today for modern reactors in terms of the control of improbable events. Experience gained from the AVR was used extensively in the design changes made to the AVR resulting in the design of the 300 MW(e) [750 MW(t)] Thorium High Temperature Reactor (THTR), which operated in Germany between 1985 and 1988. The THTR was a First-of-its-Kind production plant intended to demonstrate the viability of the different subsystem hardware designs, with specific emphasis on plant availability and maintainability. To this end, the design concentrated on building a plant with a lifetime of 40 years and an availability of 80% to 90%. Based on the experience gained from the AVR and the THTR, two German-based groups further developed pebble bed reactor designs ranging from high power reactors mainly developed by ABB (previously Brown Boveri), to the modular design of Siemens Interatom, the HTR-Modul. These two groups later combined to form Hochtemperatur Reaktorbau GmBh. In 1999, Eskom obtained the right to access the HTR engineering database that included details of the Siemens/Interatom HTR-Modul design. This design can be regarded as the forerunner of the PBMR DPP. The PBMR DPP core design was made using the same design philosophy as was used in the design of the HTR-Modul. A concept licence was issued for this reactor, and the safety arguments used in the HTR-Modul safety application are relevant for the PBMR DPP safety case. Many components used in the fuel handling and control systems of the PBMR DPP are modified copies of those used in the THTR programme. The fuel design of the PBMR DPP falls within the qualified fuel design

28 Water containing a significantly greater proportion of heavy hydrogen (deuterium) atoms to ordinary hydrogen atoms than is

found in ordinary (light) water. Heavy water is used to lower the energy of neutrons in a reactor. 29 The unit of power in the form of electricity produced by the generator.


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parameters of the German fuel programme. The actual fuel design is the same as was specified for the HTR-Modul reactor which received provisional regulatory certification in Germany. It must therefore be realised that the term First–Of–A–Kind (FOAK) used for PBMR DPP is in reality a FOAK configuration of the developed technologies namely the reactor unit and the helium turbine. Although as described above, the key components of the PBMR technology have been tested and proven, the integrated PBMR DPP is a “FOAK engineering” project. In this regard, Eskom wishes to demonstrate the techno-economic30 feasibility of the integrated system.

3.6 Functioning of the PBMR Technology

The PBMR DPP is a high-temperature, gas cooled reactor technology. It consists of a steel reactor pressure vessel, which contains and supports a metallic core barrel, which contains pebble fuel spheres. The reactor pressure vessel has a 6.2 m inner diameter and is approximately 29 m high. The annular fuel core is located in the space between central and outer graphite reflectors. Reactivity control elements can move into and out of the reactor. Two diverse reactivity control systems are provided for shutting the reactor down; one being reactivity control rods, and the other being small absorber spheres. A schematic diagram and the physical layout of the main power system are shown in Figure 3-3 and Figure 3-4 respectively.

Figure 3-3: Schematic Diagram of the PBMR Main Powe r System

30 This term refers to the combined technical and economic feasibility of the proposed configuration of the plant.


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The PBMR fuel consists of particles (TRISO coated particles) of enriched uranium dioxide (fuel kernel) coated with silicon carbide and carbon (necessary to contain the potentially harmful byproducts of the fission reaction). The particles are encased in graphite to form a fuel sphere or pebble about the size of a billiard ball. When fully loaded, the reactor will contain approximately 490 000 fuel spheres. The nuclear fission reaction within the particles encased in the fuel spheres generates heat, which is emitted into the space between the fuel pebbles in the reactor core. This heat is removed from the reactor vessel through the introduction of helium coolant that flows down between the hot fuel spheres, exiting the reactor vessel at a temperature of approximately 900 °C. The hot helium is used to drive a closed cycle gas turbine-compressor and generator system in a similar fashion as steam would drive the turbine in a coal fired power station. Please refer to Figure 3-3 and Figure 3-4.

Figure 3-4: Physical layout of the PBMR main power system

After passing through the turbine, the hot helium passes through the recuperator transferring part of the remaining heat to the gas going back to the core. A pre-cooler before the low pressure compressor and an intercooler before the high pressure compressor remove waste heat to a water based cooling system. The water in the closed circuit is cooled by chlorinated seawater through a secondary heat exchanger. At full operation, Koeberg Nuclear Power Station (KNPS) extracts 80 cubic metres (m3) of water per second from the ocean. The proposed PBMR DPP would require an additional 2,5 m3 of water per second to be extracted from the ocean. This water is chlorinated to 1 part per million (ppm) before reaching the KPNS condensers, where the water temperature increases to an average of about 10°C above ambient temperature. The seawater is proposed to be obtained from the existing Koeberg intake basin. It is proposed that the warmed and chlorinated seawater will then be returned to the sea via the existing Koeberg outfall structure. The water is to be jetted in a south-westerly direction at a speed of between 2 and 3 m/s at the outlet of the outfall structure. As the warm water is more buoyant, a warm water plume is formed.


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In the Koeberg, the surf-zone temperature standard deviation is in the order of 0.46°C. Additional water from the PBMR DPP is appr oximately 3% of the current outflow. The target markets for the PBMR technology include electric power generation and process heat applications. The 400 MW(t) module is well suited to both of the markets. For electric power generation the use of multiple units suits markets where large increments of power are not possible and allows for a staged introduction of nuclear power generating capacity.


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3.7 Demonstration Power Plant Performance and Testi ng Requirements

Eskom is seeking authorisation for the construction, commissioning, operation and decommissioning of a 400 MW(t) PBMR DPP. It is estimated that for the first seven years after construction of the PBMR DPP, operation of the facility will be primarily based on demonstrating key technical and commercial performance parameters such as construction costs, plant availability and efficiency, operational and maintenance costs and mid-life upgrade requirements. It is envisaged that after seven years of successful demonstration, the PBMR DPP will then be able to operate commercially for the remainder of its 40-year lifespan. The two main components of the PBMR DPP requiring demonstration are as follows:

• Functional integrity; and • Commercial performance.

The demonstration of the functional integrity will test the operability and the maintainability of the integrated plant system. Eskom is interested in the total plant availability, age management, online maintenance for critical equipment, and the ease of achieving the 6-yearly maintenance intervals between the general overhauls. With respect to the commercial performance of the PBMR DPP, the following aspects require demonstration:

• Direct cycle power conversion unit efficiency; • Helium leakage verification; • Operational modes and states; • Reactor unit; • Main power system; • Generator; • Maintenance procedures; • Plant availability; • Reliability; • Plant efficiency and sustainability; • Operational and maintenance cost; and • First outage.

PBMR follows an integrated Qualification Programme consisting of testing and evaluation that ensures compliance with functional performance, availability and safety requirements of the DPP. Qualification is performed for the plant functions of Structures, Systems and Components (SSC) with the following classifications:

• Safety Class High (SC-H); • First-of-a-Kind (FOAK) and Equipment with Modified Application (EMA) SSC

with Safety Class Medium (SC-M); • SC-M SSC that will directly influence an SC-H SSC in the event of the SC-M

SSC failure. Qualification is the generation and maintenance of evidence to ensure that a safety classified system and its equipment will operate within design requirements reliably on demand meeting specified safety, performance and availability functional requirements under all operating and postulated accident conditions. This includes testing or analysis or combinations thereof. The PBMR Qualification process has


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been tailored from various international nuclear plant test and qualification guidelines. It is managed for each SSC by means of a qualification programme, which is generated by following Systems Engineering process principles. The planned tests and analysis that will be required for the DPP and its systems is managed by means of a Test Evaluation Master Plan (TEMP) and the Qualification Database. A graded approach ensures that all safety classified and FOAK SSC important to safety are qualified, subjecting all SSC that will be used in the DPP to a uniform test and evaluation process approach. Qualification Requirements, considers the complete functionality of all the SSC for all the plant phases. Each SSC is treated in the same manner to identify its qualification programme, with the following inputs:

• Classification of all SSC in terms of functional description, Safety, Quality, Seismic and Environment, as well as ageing, in particular material deterioration due to radiation, e.g. reactor components;

• Detailed functions, including safety support functions, are being considered in the form of functional breakdown of the SSC at least up to sub-SSC equipment and eventually component level, linked to measurable and quantifiable characteristics determined by the envisaged plant modes and states, for Normal and Abnormal operating conditions, including postulated accident conditions; and

• Principles of functional, environmental and seismic qualification that are applied to the testing and evaluation of individual SSC prior to delivery for installation into the plant are defined in the Functional Analysis for Nuclear Safety process, the Plant Modes and States analysis, as well as in each SSC Development Specification.

During the design phase of the DPP and its SSC, preliminary qualification life cycle planning is compiled that reflects the specific off-site design verification testing and Equipment Qualification (EQ) test and analysis. Such testing must be concluded to demonstrate satisfactory functional performance of all SSC important to safety, prior to installation into the DPP. During the plant Commissioning and Plant Acceptance test phase, an Initial Testing programme is conducted on the DPP Functional Equipment groups prior to loading of nuclear fuel to demonstrate that safety systems and components perform as claimed in the Safety Case to be assessed by the National Nuclear Regulator. This is followed by the Start-up test programme after fuel load in accordance with the International Atomic Energy Association (IAEA) Safety Guide on Commissioning for Nuclear Power Plants. Such testing is included in the Plant Acceptance Test plan. Although the DPP technology is based on the proven German Thorium High-temperature Reactor (THTR) design, it may be regarded as a FOAK plant due to application of the direct Brayton-cycle amongst others, such as the new core design with solid center reflector and some passive engineering features. Where the qualification of nuclear power plants traditionally only deals with the qualification of SSC required for the safety case, FOAK equipment with SC-H and the commissioning of the installed plant, the DPP also has an element of plant qualification. With qualification of SSC (or Equipment Qualification) is meant the qualification of individual SSC that is performed external to the plant (i.e. Off-site test facilities dedicated for testing of the SSC, for example performed at the HTF). Plant qualification means the validation and acceptance testing of the integrated plant Functional Equipment Groups (FEG). A dual Qualification process, comprising both an Equipment Qualification programme and a Plant Acceptance Test programme,


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which includes initial testing, is therefore proposed. This programme is referred to as the DPP integrated qualification programme, which is outlined in Figure 3-5 . The Equipment Qualification (EQ) process consists of the EQ Verification Phase, EQ Validation Phase and EQ Preservation Phase. The Pre-delivery testing phase as part of EQ Verification includes all the Off-site design verification and Equipment Qualification testing and analysis for extreme load case conditions, prior to installation of SSC into the plant. The Plant Acceptance Test Phase includes the Installation Inspection activities, Site-based Acceptance Testing of SSC, followed by Initial Testing of integrated systems. This may also be termed as validation of plant FEG. Initial Testing is divided into Pre-Fuel load testing of the DPP systems and Post-Fuel load Start-up Testing. The site-based acceptance test programme therefore interfaces with the License Variation to load nuclear fuel. Operational testing follows Hand-over of the DPP to Eskom as the operating utility group and includes the Plant Acceptance Testing, Operational Testing and Performance Monitoring test phases. Plant Acceptance testing (PAT) verifies that the integrated plant complies with contractual requirements, prior to handover. The tests conducted during this stage verify performance of the plant against contractual requirements and will not go beyond the normal operational envelope of the plant. Operational Testing (OT) includes all IST, ISI, HCM and Continuous Plant Operations Monitoring (OPM) conducted by the operations staff throughout the service life of the plant to ensure that all systems are functioning to specification, and that the qualification status is maintained. In Service Inspection (ISI), In Service Testing (IST), Operating Performance Measurement (OPM) and Health Condition Monitoring (HCM) tests identified during the qualification planning will be included in the TSP. Performance Monitoring (PM) is the engineering activity concerned with the validation of the SSC and the plant’s performance against their development specifications. Engineering data is gathered by PBMR engineers from operational tests and monitoring instrumentation in order to validate and update the plant and SSC design data packs, software and computer models. PBMR is pro-actively developing qualification plans of the various SSC of the DPP qualification plans in parallel with the Design process, with the current focus on generation of Design Verification and Equipment Qualification tests for many SSC. The objective is to have all main SSC fully tested prior to installation into the DPP plant, where after integration testing and Site-acceptance testing will follow in a progressive manner.

ESKOM PBMR DPP EIA Version 1.0 / October 2009 Final Plan of Study for EIA

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Figure 3-5: PBMR DPP Integrated Qualification Proce ss

ESKOM PBMR DPP EIA Final Plan of Study for EIA

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3.8 Construction Phase Infrastructure Requirements

Infrastructure required to be constructed and/or altered during the construction phase of the development consists of the following:

3.8.1 General Infrastructure This includes infrastructure such as repair shops, warehouses and parking areas for construction vehicles and machinery. The proposed layout of these facilities is indicated on the excavation layout plan in Appendix F .

3.8.2 Proposed Main Construction Services The main construction services include a potable water supply pipeline, power supply and data cables. The potable water supply pipeline is anticipated to run from the municipal bulk water main (running parallel to the R27) to the PBMR DPP site via the existing firebreak that runs adjacent to the Eskom HV power lines. The pipe work for the potable water will be 250 mm not bulged (Normal Bore) high density polyethylene (HDPE) piping. The maximum flow rate through the line will be 120m3/hr for a period of approximately 4 months. The water usage before and after this peak period will be on average 60% of the peak demand. The water pressure will be between 6 and 8 bar G. The dimensions of the trench will be approximately 1,2m depth by 1m width. In order to gain access for manpower and equipment, a five (5) metre servitude will be required, which will be restored as far as possible after construction. Approximately four (4) valve boxes will be introduced to allow future access to the pipeline. The power supply for the R27 construction yard will be sourced via 2 x 11kV cables from a switchyard that is to be erected by Eskom adjacent to the KNPS bulk stores, and routed directly across the road (R27 to ACP) to join up with the potable water servitude. Optical fibre cables will be routed from a point between the bulk stores and the environmental laboratories. The cable is to be routed directly underneath the road from the R27 to ACP and will join up with the potable water servitude. The total servitude will be jacked underneath the R27 by an experienced and professional pipe jacking company. This process will be approved by the Department of National Roads. The associated satellite photo shows the routing of the three utilities in colour coding (Figure 3-6 ).


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Figure 3-6: Routing of the three utilities Contractor Yard (Laydown Areas) Areas for the laydown of construction materials and equipment are required during the construction phase. Two alternative areas have been identified for assessment. (See section 5 of the EIR for details in this regard). Temporary Bypass around the Modder River Bridge

During the transportation of heavy equipment from Saldanha Bay Harbour to the PBMR DPP Site, it will be necessary to cross the Modder River (see section 3.8). To avoid damaging the Modder River Bridge, which is located approximately 27 km from the R27 turn-off to the KNPS site, it may be necessary to construct a temporary bypass on the downstream side (Figure 3-7) of the Modder River Bridge. The bypass would be approximately 11 m wide and 250 m in length with pipes placed to take the stream flow. It has been recommended that the bypass be constructed during the dry season and then used as required. It is important to note that this bypass will only be needed if further investigations suggest that it will not be possible to cross the bridge using an appropriate engineering solution. In this regard, the feasibility of using temporary beams supported by cranes upon which the reactor pressure vessel will slide across the bridge is under consideration. The specialist studies in wetland ecology, flora and fauna have considered the impacts associated with the temporary bypass in the event that it becomes necessary.


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Figure 3-7: Downstream side of the Modder River Bri dge.

3.8.3 Widening of Intersections Temporary civil works will be required to widen the intersections of the R559 and the R79 as well as at the intersection of the R79 with the R27. This widening is necessary to allow for the width (approximately 8 m) of the combination trailer transporting the reactor pressure vessel from Saldanha Bay Harbour to the PBMR DPP Site.

3.8.4 Avoidance of Telkom Lines and Power Lines Approximately 10 Telkom line crossings have been identified along the route proposed for the transportation of heavy equipment from Saldanha Bay Harbour to the PBMR DPP Site. The lines will need to be cut and replaced after the load has moved through. A cutting will be constructed parallel to the R27 to allow oversized loads sufficient clearance beneath the Eskom HV power lines.

3.8.5 Propping of Culverts Large culverts along the roads proposed for the transport of the heavy equipment will need to be temporarily propped in order to prevent structural damage.

3.8.6 Improvement of Picnic Areas for Laybye Areas Six picnic spots have been identified along the proposed route for transportation of the heavy equipment from Saldanha Bay. It is proposed that these picnic spots be upgraded for use as a traffic laybye.

3.8.7 Accommodation for Construction Workers Accommodation for approximately 800 workers will also be required during the construction phase. Initially, this requirement had been included in the scope of the EIA study, however, it was later proposed that existing accommodation in surrounding areas be utilised for construction workers and therefore no construction village will be required.


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It is anticipated that approximately 70% of the labour force will be sourced from the local communities which will reduce the need for additional accommodation. It is also anticipated that the remaining labour force (which is anticipated to be skilled labour) could be accommodated in existing tourism infrastructure. The promotion of sourcing local labour will also reduce the impact and pressure on service delivery for the City of Cape Town as well as reduce potential impacts on land use, surrounding communities, health and other socio-economic factors such as the introduction of people dissimilar in profile and the inflow of workers from outside the local area. Based on the above, the need for the additional construction village was deemed unnecessary and has been removed from the scope of this EIA.

3.9 Transportation of Heavy Equipment to the Propos ed Site

Preliminary investigations have been undertaken by PBMR (Pty) Ltd. with respect to the transportation of the exceptionally heavy components of the PBMR DPP from Saldanha Bay to the KNPS site. Exceptionally heavy components, such as the reactor pressure vessel are likely to have a mass of more than 900 tonnes. To transport this load, it is currently proposed that a combination trailer, consisting of 3 trailers, each with a length of 42 m and a combined width of approximately 8 m be used to reduce the load on the road surface. It is anticipated that a specialist logistics company will be contracted to provide the necessary route study, to obtain approvals and to provide the method of transportation. The load is likely to be delivered by ship to the Saldanha Bay Harbour where it will be offloaded onto the proposed combination trailer. Using existing infrastructure as far as possible and with some minor upgrading of this infrastructure (as detailed in section 3.6 ), the load will make use of the R559 travelling east (Figure 3-8) . At the intersection with the R559 and the R79, the load will turn left and proceed to the intersection of the R79 with the R27. The load for the reactor pressure vessel will travel at approximately 4 to 5 km/hour (approximately 3 days travel time) and will take up the entire width of the road at times. This will have a significant impact on traffic flow during transportation, especially on the R27. Traffic laybye areas (existing picnic areas which will be upgraded) have been identified along the route to allow for the load to move off the road so that traffic can pass.


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Figure 3-8: Possible route from Saldanha Harbour to the R559

3.10 Operational Phase Infrastructure Requirements

The following infrastructure will be required for the duration of the operational phase of the proposed PBMR DPP (please refer to Appendix E for a layout plan and approximate dimensions of the infrastructure):

3.10.1 Module Building The nuclear reactor and associated components are housed in the reactor building. The reactor building structure is comprised of reinforced concrete. The foundation for the building comprises of an approximately 3 m thick raft, founded on bedrock approximately 26 m below surface level. The surface level around the reactor building at the proposed site is at an elevation of approximately 13.5 m above mean sea level (amsl). The reactor building will have a height of approximately 60 m above surface level. A chimney stack of approximately 15 m in height will extend from the reactor building. Generator and Main electrical power system The generator and associated electrical and auxillary power plant are located in a generator building which is situated adjacent to the northern gable of the reactor building. The generator house comprises of a conventional framed structure,


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constructed of reinforced concrete to 3 m above the generator floor, located approximately 10 m above surface. Above this level a structural steel support system, covered with aluminium sheeting, is proposed.

3.10.2 Radioactive Waste Handling Building The services building houses the main control room and the waste handling and storage system and also provides the controlled access to the reactor building.

3.10.3 Standby Diesel Generator Building The ancillary building is located to the east of the reactor building and north of the services building and houses the medium and low voltage switchgear, the diesel generators, and other systems associated with the operation of the PBMR DPP. Underground tunnels interconnect the reactor building with the services and ancillary buildings.

3.10.4 Cooling Water Plant Building The helium gas that cycles through the reactor and drives the turbine is indirectly cooled with sea water. A cooling water plant building is located to the west of the generator building and houses the cooling water pumps and heat exchangers. Piping between the cooling water plant building and the reactor building is routed via an underground tunnel.

3.10.5 Administration Office Building An administration office building on the south west corner of the terrace will house the PBMR DPP staff. The services building, ancillary building, administration building and cooling water plant building are likely to be constructed using conventional beam column frames supporting reinforced concrete floors and structural steel clad roofs.

3.10.6 132 kV Transmission Power Line and Extension of the Duine Substation A 132 kV transmission power line (double circuit) is proposed to be constructed from the proposed PBMR DPP to the existing Duine Substation and the Koeberg Substation (Figure 3-9) . This transmission line links the proposed PBMR DPP to the national transmission network. The type of towers to be used will be either round steel poles or lattice (Figure 3-10) . Each tower will be approximately 31 m in height. The width of the servitude likely to be required is between 30 m and 40 m. To accommodate the 132 kV power line, the existing Duine Substation will need to be increased by 25 m towards the north and by 45 m towards the south.


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Figure 3-9: Proposed alignment of the 132 kV power line. The red dotted line is the proposed 132 kV power line. All other colou red lines are existing power lines

Figure 3-10: Round steel pole tower structure.

Figure 3-11: Lattice tower structure


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3.10.7 Internal and External Roads Widening of a portion of the road to the KNPS from the R27 turnoff and the construction of internal roads for access to the PBMR DPP site is also proposed.

3.10.8 Existing Infrastructure on the KNPS Site Existing infrastructure on the KNPS site is proposed to be used as far as possible for the PBMR DPP. This infrastructure includes the following:

• Cooling water from the sea using the existing intake basin and outflow structures;

• Low and intermediate level radioactive waste management and storage structures and systems for the processing of such waste prior to disposal to Vaalputs Waste Disposal Site in the Northern Cape;

• Existing transmission power line network including substations; • Sewage treatment facilities; • Certain internal roads; and • Existing security measures.

3.11 Transportation of fuel to the proposed site In 2001-2002, an EIA was undertaken for the proposed manufacturing of nuclear fuel at Pelindaba in the North-West Province for the proposed PBMR, and the associated transportation of nuclear materials. The proposed project comprised the following31:

• The establishment, operation and decommissioning of a fuel manufacturing plant within the BEVA complex at Pelindaba. The fuel would be for specific use at a Demonstration Module PBMR.

• Transportation of imported enriched uranium from the port of entry (Durban) to Pelindaba, and of manufactured fuel from Pelindaba to the proposed Demonstration module PBMR at the preferred site of Koeberg in the Western Cape.

The EIA process for fuel manufacturing plant identified issues to be explored in the EIR which can be subdivided into two broad categories, namely:

• Fuel manufacturing o Technical (biophysical or engineering); o Waste (radiological and non-radiological); o Economic/Financial; o Safety/Health and Security; o Spatial planning/Land use; o Institutional capacities; and o Socio-economic.

31 Summarised from the Executive Summary of the Final Impact Report for the Proposed Manufacturing of Nuclear Fuel at Pelindaba in the North West Province for the Proposed Pebble Bed Modular Reactor and Associated Transportation of Nuclear Materials from Durban to Pelindaba, as well as Manufactured fuel (Pebbles) from Pelindaba to the Preferred PBMR Site at the Koeberg Nuclear Power Station in the Western Cape, October 2002, undertaken by the PBMR EIA Consortium, as included in Appendix AR


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• Transportation of nuclear material o Technical; o Waste; o Socio-economic; o Safety/Health; and o Spatial use/Institutional capacities

During the impact assessment, the major risks were identified these include:

• Product – Handling and Storage – Fuel spheres stolen of sabotaged; • Product – Handling and Storage – Fires occurring; • Process Plant – Fuel manufacturing – Fires occurring; and • Process Plant – Fuel manufacturing – Non-natural events.

The EIR executive summary (Appendix AR) describes the above issues in greater detail): Based on the risk assessment undertaken, the above issues are all related to human factor aspects. Therefore, a key conclusion drawn was that management systems must be put in place for all the aspects related to the loss of raw material and fuel spheres; systems and equipment to deal with the external effects of fires; and to assure that the design for external impacts is sound. In terms of transportation of fuel to the proposed PBMR site, the EIA showed that:

• There is no generic advantage of road over rail as the preferred mode of transport;

• Road is the preferable mode of transport due to more limited handling, low volumes and low frequency of movement of materials;

• Radiological impacts from transport are low, due to package design and safety controls; and

• In the event of an in-transit spill of nuclear material, limited impact on the environment is foreseen, due to mass and low radioactivity of the nuclear material.

The EIR therefore recommended that the proposed plant and transportation of nuclear material be authorised as there is no significant environmental risk, provided that:

• The proposed PBMR DPP is authorised; • The Environmental Management Plans and recommendations contained in

the EIR are implemented; • The framework Transport Plan is developed into an agreed final Transport

Plan in conjunction with the authorities; • The recommendations of the Social Impact Assessment are formalised and

implemented; • Extensive environmental monitoring be conducted as recommended

(radiological/non-radiological); and • Financial provision be made for decommissioning and long-term storage of

radioactive waste. After the consideration of a variety of appeals, the DEA upheld the positive record of decision for this EIA, with detailed conditions (refer to Appendix AS ).


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3.12 Nuclear Fuel Cycle of the PBMR DPP

3.12.1 Introduction 32

Internationally, a nuclear fuel cycle is recognised. It starts with mining and moves through the extraction of the uranium, and the conversion of the uranium into a feed material for uranium enrichment. After uranium enrichment, nuclear fuel is fabricated, which means that the uranium is built into a structural assembly that is suitable for insertion into the particular type of nuclear reactor under consideration. When the fuel is ‘spent’, which means that all or much of the uranium has been used up, the fuel elements are removed and then sent to spent fuel storage. At this point there are time options, if the fuel is highly depleted it will be sent to a waste repository. Each of the stages in the nuclear fuel cycle, as it will relate to the proposed PBMR DPP are explained below.

3.12.2 Mining and Milling 33 Mining and milling represent the first two processes in the nuclear fuel cycle. Uranium is an ore that is mined either as open cut or underground mining, or through a process of in situ leaching, which is dependent on the depth at which the ore may be found. Once mined the ore is sent to a mill where it is crushed and ground to a slurry. Sulphuric acid is then added to dissolve the uranium from the other materials. The uranium solution is then filtered, separated and dried to produce a solid uranium concentrate termed “yellow cake”. It is this uranium oxide concentrate or U3O8 that is used in the next step of the cycle. Approximately 200 tonnes of U3O8 is required on an annual basis to maintain a 1 000 MW nuclear power reactor. If compared to oil or coal, 1 800 000 tonnes of oil and 3 000 000 tonnes of coal would be required to produce the same amount of electricity.

3.12.3 Enrichment Atoms with the same number of protons but with different numbers of neutrons are called isotopes of the original atom. Chemically these substances behave the same but they have slightly different physical properties. Uranium has two isotopes that are of interest, one is Uranium (U)-238 and the other is U-235. When uranium is mined, it comes out of the ground as 99,3% of the U-238 and only 0,7% of the U-235. The latter, because it is easier to split, is the isotope that is required for nuclear energy use (Appendix D) . Enrichment refers to the process of concentrating the amount of Uranium (U)-235 isotope, compared with the U-238 isotope. Enriched uranium for the PBMR DPP will be delivered to the proposed fuel plant at Pelindaba in the form of U3O8, which is powder. The level of enrichment will be below 10%. This is called “Low Enriched Uranium”, or LEU.

32 Appendix D 33 Arcus GIBB (2008)


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There will be approximately 3 tons of uranium in the core of the PBMR DPP, and approximately 1 ton of uranium will be replaced with fresh fuel per year of operation. The LEU will be bought from a foreign supplier and transported to Pelindaba in special transport containers34.

3.12.4 Fuel Fabrication The fuel fabrication process will take place at the Pelindaba site of the Nuclear Energy Corporation of South Africa (Necsa) near Pretoria. The fabricated fuel for the PBMR DPP consists of graphite balls about the size of tennis balls. Each ball contains 9 grams of enriched uranium in the form of grains about the size of grains of sugar. These grains are mixed evenly into the graphite ball much like rolling baking dough and sugar grains together (Appendix D) . Each spherical PBMR DPP fuel pebble is made from matrix graphite, which is a mixture of natural graphite, electrographite, and a phenolic resin that acts as a binder. It consists of an inner region that contains fuel in the form of spherical coated particles embedded in the graphite matrix. Each coated particle consists of a spherical uranium dioxide kernel surrounded by four concentric layers. The first layer surrounding the kernel is a porous pyrocarbon layer, known as the buffer layer. This is followed by an inner high-density pyrocarbon layer, a silicon carbide layer, and an outer high-density pyrocarbon layer. The coated particles are embedded in a graphite fuel sphere (Figure 3-12) . The function of the matrix graphite is to form the main structure of the fuel sphere and to contain the coated particles and to provide a heat conduction path between the coated particles and the reactor coolant. The matrix graphite also acts as the moderator for neutrons in the PBMR DPP core. The reactor core will contain approximately 452 000 fuel spheres when fully loaded.

Figure 3-12: Fuel sphere design for the PBMR DPP

34 PBMR EIA Consortium (2002)


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3.12.5 Spent Fuel Storage The term “spent fuel” are used to denote fuel that has been removed from a nuclear reactor when its useful life is complete (Appendix D) . However, depending on the type of reactor, the actual uranium, can still be used to various degrees. The PBMR DPP will make use of online fuelling which means that the reactor will not be taken out of service during refuelling. The fuel is introduced at the top of the reactor while used fuel is removed at the bottom to keep the reactor operating at full power (Figure 3-13) .

Figure 3-13: Schematic layout of the fuel handling and storage system Fuel pebbles continuously move through the core from the top to the bottom during the normal operation of the plant. Once a pebble exits the bottom of the reactor it is measured and tested to ensure that it conforms to the physical integrity specifications. It is also evaluated for burn-up. Pebbles that are physically in good order and have not reached the target burn-up are returned to the top of the reactor for re-introduction into the core.


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Rejected pebbles are transferred to the spent fuel tanks in the basement of the plant for storage. 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 spent fuel tanks are designed to have sufficient capacity to store the full spent inventory of the expected 40 year life cycle of the PBMR DPP. The fuel storage is integrated into the confinement and chimneys are included. The radioactivity of the spent fuel results in heating in the tanks. Thermal cooling of the spent fuel storage vessels is thus required and this will be in a water pool followed by passive dry cooling and a naturally ventilated chimney system. To deal with the PBMR DPP spent fuel, the following options are being considered at present (PBMR, 2005)12:

• The direct disposal route implies storage of spent fuel for up to another 40 years after final shutdown of the plant, after which it will be transferred to shipping/storing casks for shipment and disposal at a designated high-level waste repository. and

• Consideration is also given to reduction of the minor actinides in the fuel. Such a reduction programme will require closing the fuel cycle, with fuel reprocessing as a necessary intermediate step. Reprocessing of fuel (and therefore minor actinide reduction) will have to be seen as part of future international collaboration on fuel cycle matters.

3.12.6 Reprocessing 35

Reprocessing is a chemical process to separate any usable elements (e.g. uranium and plutonium) from fission products and other materials in spent fuels. Usually the goal is to recycle the reprocessed uranium or place these elements in new mixed oxide fuel (MOX). While reprocessing of spent fuel is not excluded as an option for spent fuel management in the National Radioactive Waste Management Policy and Strategy (2005), there is no intention to reprocess the PBMR DPP spent fuel at present.

3.12.7 Final Disposal Final disposition of waste is understood to refer to final disposal. Other options, such as long-term storage, do not represent final disposition. At some time in the future, all such alternative options must end in final disposal, for only disposal meets the fundamental principles of radioactive waste management in the long term. There is currently no final disposal solution for high-level radioactive waste in South Africa. Consequently, it is proposed that all high-level waste generated by the PBMR DPP will be stored within the reactor building of the facility according to National Nuclear Regulator (NNR) requirements.

35 Appendix D


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3.13 Nuclear Licensing of the PBMR DPP

3.13.1 Obtaining a Nuclear Licence in South Africa

The NNR is mandated by the National Nuclear Regulator Act 1999, (Act 47 of 1999, “the NNRA”) to provide for the protection of persons, property and the environment against nuclear damage through the establishment of safety standards and regulatory practices. These practices include the regulatory control over the siting, design, construction, operation, manufacture of component parts, decontamination and decommissioning of a nuclear station through the granting of a nuclear licence. Assurance of compliance to the conditions of the license is obtained through the implementation of a series of compliance inspections. For a utility (such as Eskom) to obtain a licence to site, construct and operate a NPS, like the PBMR DPP, in South Africa it must follow a lengthy licensing process. This includes extensive technical analyses to demonstrate compliance to the NNR’s safety standards. In accordance with Section 21 of the NNRA, Eskom has submitted a formal application to the NNR for a nuclear installation license for the siting, construction, operation, decontamination and decommissioning of the proposed PBMR DPP. The application contains a description of the nuclear installation and the proposed location. Comprehensive technical submissions have and will continue to be submitted to the NNR demonstrating that the proposed plant meets all the NNR’s safety standards. The NNR will review and analyse the technical submissions and then make a recommendation to the NNR Board as to whether or not to issue a nuclear installation licence. The NNR Board must be satisfied that all relevant concerns and questions with regard to radiological safety have been appropriately considered by the NNR during the technical review. The NNRA makes provision for any person, who may be directly affected by the granting of a nuclear installation license to make representations to the NNR Board, relating to health, safety and environmental issues connected with the application for the license. The NNRA also makes provision for the NNR Board to hold public hearings on an application for a nuclear installation license. A staged licensing process is normally followed. For each stage (such as bringing nuclear fuel onto the site, loading fuel into the reactor, taking the reactor critical), the licence will contain license conditions, which will restrict what can be undertaken at that stage, thereby allowing the NNR to exercise regulatory control. Only when defined pre-requisites are met and the NNR is satisfied, will the licence be “varied” to allow the next stage activities to proceed. The regulatory process does not end when an operating licence is issued. Throughout the construction and operation of a plant, the NNR conducts inspections to ensure strict compliance with licensing conditions. The NNR also requires the utility such as Eskom to implement a programme of inspections to ensure compliance with all conditions of the nuclear licence. If there are any violations, the NNR can take legal action against the utility, which could include revoking its licence. In July 2000 Eskom applied to the NNR for a nuclear installation licence for the prospective siting, construction, operation, decontamination and decommissioning of


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a demonstration unit of a 110 MWe Class PBMR electricity generating power station. This was based on the expected design output at the time. Since 2000 the design output has been increased and thus an amended license application is required. This application amendment will occur when the final safety case submission for the construction license is submitted.

3.13.2 Basic Licensing Requirements The principles that must be met to ensure safety in any nuclear installation in South Africa are presented in the Regulations on Safety Standards and Regulatory Practices (RSSRP) published as Regulation No. R388 dated 28 April 2006, promulgated in terms of the NNRA. The NNR published Basic Licensing Requirements (BLR) for the PBMR DPP, in a Requirements Document (RD-0018) in 2007. The BLR for the PBMR DPP are based on and are established to fulfil the principles of the NNRA including the RSSRP. The BLR for the PBMR DPP indicates that a licence is required for each of the stages involved in a nuclear installation namely siting, construction, operation, decontamination and decommissioning of the facility. Stages may be combined into a single licence application only if the applicant is able to provide appropriate safety documentation for all proposed stages through a safety case. Some specific Principal Safety Requirements for the PBMR DPP which the BLR require are as follows:

• The facility must be designed, constructed, commissioned, operated, maintained and decommissioned according to a standard which meets good engineering practice;

• In line with the Principal Safety Requirements of [1], the principles of Defence-in-Depth (DiD) must be applied to the PBMR in a manner consistent with the DiD processes described in the appropriate international safety standards and related documents (e.g. Safety Reports produced by the IAEA) so that there are multiple layers of PBMR Functions provided by the Structures, Systems and Components (SSC), and procedures, (or a combination thereof) to ensure that the Fundamental Safety Functions (FSF) of Heat Removal / Reactivity Control /Confinement of Radioactivity are met. Event prevention and event mitigation are natural consequences of the DiD principle. The application of the DiD Principle to the design and operation of the PBMR is elaborated further in Appendix C of the BLR for the PBMR DPP;

• In line with the Principal Safety Requirements of the RSSRP, the ALARA (As Low As Reasonably Achievable) principle must be adopted in a manner consistent with the ALARA processes described in the appropriate international guides (e.g. reports produced by the ICRP.) The application of the ALARA principle is required for selection of design and operational features that provide an adequate minimisation of radiological doses and thus the optimum level of safety in terms of radiological risks. The application of the ALARA principle to the design and operation of PBMR is elaborated further in Appendix D of the BLR for the PBMR DPP; and

• The facility design and its proposed operation must meet the dose and risk limits as defined in Annexure 2 and 3 of the RSSRP respectively.


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3.14 Safety of Nuclear Power Stations

3.14.1 General Safety Considerations 36

Since the commercial production of nuclear energy to generate electricity began, it has arguably proved to be one of the world’s safest energy technologies. This may in part be contributed to the fact that safety forms a major component of the design, construction, and by operation of a NPS. There are a number of systems that monitor, control, and support the safe operation of the reactor at each power plant. These systems provide maximum safety and reliability and reduce the chance of an accidental release of radioactivity into the environment. There are numerous safety systems that have been engineered to assist in preventing an accident with the reactor or to lessen the effects in the event that an accident should occur. All critical safety systems have backup systems that duplicate the jobs that the system is supposed to perform. Another key aspect to consider when looking into the safety of a NPS is the training and preparedness to which the people who operate these stations are exposed. For example, reactor operators are trained and tested on the procedures of power plant operation, and in order to train such staff, utilities around the world use sophisticated power plant simulators, which are replicas of the control room of the real power plant in which they will be working. The simulators are computer controlled, allowing the operators to gain practical experience in managing all types of normal and unusual occurrences without posing any danger to the public or the environment. The nuclear industry throughout the world has rigid safety standards. In South Africa these standards are set and regulated by the NNR, and Eskom would have to prove to the NNR that the proposed PBMR DPP can and will meet these stringent safety standards before it can be built and operated. Periodic inspections also ensure that each facility operates safely. It is a misconception amongst the general public that a nuclear reactor can explode as an atomic bomb. This cannot happen, as a nuclear explosion requires a very high concentration of fissionable uranium, which is not the type of uranium that is used within the PBMR DPP. The uranium used in the PBMR DPP is enriched to a maximum of about 10 %, whereas an atomic bomb using uranium is enriched to above 90 %.

3.14.2 Historical Nuclear Accidents 16 The International Atomic Energy Agency (IAEA) has developed a scale called the International Nuclear Event Scale (INES), which they use to rate the severity of a nuclear accident on a scale from 0 to 7. Two nuclear accidents for which they have applied this scale are Three Mile Island (located on an island in the Susquehanna River 10 miles from the Pennsylvania capital of Harrisburg in the United States of America) and the Chernobyl disaster (in the former Soviet Union near that present day Belarus – Ukraine Border).

36 Adapted from Arcus GIBB (2008)


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(a) Three Mile Island (1979) Considered to be the worst civilian nuclear accident outside of the Soviet Union (INES rating of level 5). The reactor of Unit 2 experienced a partial core meltdown, but the reactor vessel and containment building were not breached. Only very small amounts of radiation were released to the environment. This accident resulted in a number of changes to the manner in which NPS are operated and maintained in the West.

(b) Chernobyl (1986) The worst nuclear accident in history and the only event to date that has received an INES rating of level 7. This accident spread radioactive contamination across large portions of Europe. RMBK reactors can be safely operated despite the identified design flaws, the cause of this accident can be, in part, attributed to a disregard for procedures and the bypassing of safety systems by the operators. Significant design changes have now been pursued to prevent a reoccurrence of such a disaster in the future.

3.14.3 Deaths related to various forms of energy pr oduction 37 Many occupational accident statistics were generated over the last 40 years with respect to nuclear reactor operations in the US and UK. When comparing the data with that of coal-fired power generation, it is evident that nuclear is a distinctly safer method of producing electricity. A major reason for coal's unfavourable showing is the huge amount of ore, which must be mined and transported to supply even a single large power station. Mining and multiple handling of large quantities of material of any kind involves hazards as reflected in the statistics presented below, which show that nuclear power, if it is continued to be managed correctly, is one of the safest forms of power generation available. A comparison of deaths per TWy (Terra Watt year)38 as a result of accidents among different forms of energy production was undertaken by the World Nuclear Association. The results of this study are as follows:

Hydropower - 885 deaths; Coal - 342 deaths; Natural Gas - 85 deaths; and Nuclear - 8 deaths.

3.14.4 Safety features of the PBMR DPP

The PBMR DPP includes in its design multiple passive39 features, based on laws of nature, that will by themselves prevent any mishap from progressing to a substantial release of radioactivity or to the equivalent of a core melt accident (pers. comm. PBMR Client Office, Eskom, 2009). RD-0018 (see section 3.10.2 above) gives maximum allowable doses to the public for

37 Adapted from Arcus GIBB (2008) 38Basis: per million MW operating for one year, not including plant construction, based on historic data, which is unlikely to

represent current safety levels in any of the industries concerned. 39 Please note that the term inherently safe has been replaced in the Final EIR with the term multiple passive features. This is to expand the clarity on safety means in relation to the PBMR DPP design. Multiple passive features, as opposed to inherently safe, amplifies the fact that the PBMR DPP has multiple safety features based on the laws of nature, that will by themselves prevent any mishap from progressing to a substantial release of radioactivity or to the equivalent of a core melt accident (pers. comm. PBMR Client Office, Eskom, 2009). When reading the specialist studies, please understand the term inherently safe to mean multiple passive features.


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events with different frequencies. It also requires application of ALARA and has a number of other general requirements. Some licensing guides and directives are also issued dealing with core design, graphite material, quality assurance etc. All these are used as design requirements for the PBMR DPP. However, beyond that PBMR (Pty) Ltd. has formulated a number of additional goals to be achieved in the design. These can be summarised as follows40:

The design for the PBMR DPP must ensure that significant fuel damage is impossible under any conditions that can be predicted;

Response to accident conditions must not rely on active systems to meet the requirements in the bullet point above. This includes having to add primary coolant or repair the pressure boundary;

There shall be no need for operator actions to meet requirements in the first bullet point above for the first 24 hours. (This does exclude actions to minimise the damage to the plant and/or the release of activity in that period).

The nuclear accidents mentioned in section 3.11.2 are not considered to be possible for the PBMR DPP. This is because of the following attributes of the PBMR DPP technology41:

The removal of decay heat is independent of the reactor coolant conditions; 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 peak temperature that can be reached in the core of the reactor (1600 °C) is well below the temperature that may cause damage to the fuel;

The PBMR has a strong negative temperature coefficient of reactivity which stops the nuclear chain reaction and results in the reactor shutting down by itself;

Analysis has shown that a reactor core meltdown for the PBMR DPP is not credible;

The reactor is housed in a building, part of which is a strengthened enclosure around the Main Power System (MPS). The building is designed to withstand significant external forces such as aircraft impacts, seismic events, tornadoes or explosions caused by saboteurs; and

The walls surrounding the reactor pressure vessel are 2.2 m thick reinforced concrete.

3.15 Decommissioning of the PBMR DPP

A decommissioning plan for the PBMR DPP has been compiled by PBMR (Pty) Ltd. The plan (PBMR, 2007) recognises that appropriate planning for decommissioning must be addressed at the design stage. The following decommissioning options are identified within the plan for the PBMR DPP:

• Transferring the plant into a safe state, i.e. conversion into a dormant state which forms a safe enclosure for a long period of time to benefit from decay of

40 PBMR (2007) 41 Summarised from section 4.5 of the Revised Final Environmental Scoping Report


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radioactive materials over time, and decommission the plant at the end of this period; or

• Converting the plant into another facility which, in part, makes use of the existing Structures, Systems and Components (SSC); or

• Dismantling the plant in totality directly after shutdown to prepare the site for other uses.

The decision as to which option will be pursued will, according to the plan, be determined towards the end of the operational phase. The decommissioning scheme to be adopted for the PBMR, i.e. early plant dismantlement or safe enclosure, will be determined by the nuclear strategy regarding the back-end fuel cycle, and waste disposal methodology at the time of decommissioning. The choice is between the following options:

• Decommissioning directly after plant shutdown; • Decommissioning after an undefined period of safe enclosure; and • Decommissioning after 40 years of safe enclosure.

Of these options, one scenario is chosen as a representative sample and the decommissioning strategy is developed along the lines of this scenario. The option chosen is the decommissioning after 40 years safe enclosure. The reasons for choosing this scenario are:

• This scenario is in agreement with the design philosophy of the plant; • The ALARA principle is fully adhered to by making maximum use of

radioactive decay; and • A large part of the waste will have decayed to low level, but long-lived waste.

Options also exist as to the method of decommissioning. The main options are between:

• Removal of the Reactor Pressure Vessel (RPV) as a unit together with all internals (except fuel and control rods);

• Removal of the graphite and all other RPV internals for radioactive waste disposal; and

• Subsequent cutting the RPV into manageable sections. The second option is chosen because:

• The building provides containment and protection against releases of radioactivity right up to the completion of the removal of all radioactivity from the facility;

• Although there is an increase in volume of waste by removing the graphite from the RPV (where it is in a minimum volume configuration), the increase in the total volume is not drastic; and

• The road transport of an approximately 2 000 ton load to currently the only nuclear waste repository, Vaalputs, would constitute a problem at the present time.

It is a requirement of the NNRA that a decommissioning fund be established by the holder of a nuclear licence to ensure that there are sufficient funds available for decommissioning of the facility. In addition to this requirement, PBMR (Pty) Ltd. has


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also agreed to make financial provision for the premature decommissioning of the PBMR DPP, should this be necessary. It can be noted that, in terms of the EIA Regulations promulgated under the National Environmental Management Act, 1998 (Act 107 of 1998), a separate environmental authorisation will be required for the decommissioning of the PBMR DPP.

3.16 Technical Specifications of the PBMR DPP

Selected technical specifications of the PBMR DPP upon which this EIA has been based are provided in Table 3-2 .

Table 3-2: Selected Technical Specifications of the PBMR DPP

Design Parameters Description(s) Comment

PBMR DPP Plant

Location X: 3 728 389

Y: - 52 239

WGS Coordinates of approximate centre point of facility. The plant is approximately 400 m south of the existing KNPS

PBMR Reactor Building (Nuclear Island) (L x W x H)

78 m x 60 m x 55 m Actual dimensions may be 10 m more or less than those provided

Generator House (Conventional Island)

(L x W x H)

28 m x 37 m x 37 m Actual dimensions may be 10 m more or less than those provided

Emergency Planning Zone 5 km As per existing 5 km EPZ for the KNPS

Construction

Reactor Building

Robust protective enclosure with controllable radionuclide retention function

Safe Shutdown Earthquake The plant is designed for a safe shutdown for the seismic conditions appropriate to the Koeberg site

Power

Total Thermal power (Pn) Nominal 400 MW(t)

Maximum Continuous Rating (MCR)

Nominal 165 MW(e)


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Power conversion Single-shaft Brayton cycle with helium as coolant

Shaft is horizontal

Generator placement In conventional island

Core

Core shape Annular cylinder around near-solid central graphite reflector

Fuel

Fuel type TRISO coated Uranium dioxide particle

Fuel to be manufactured at the Pelindaba facility

Fuel enrichment

start up core

normal operation

4.5%

9.6%

Fuel configuration Coated particles in 60 mm diameter graphite spheres

Uranium content per sphere 9 g Uranium per sphere

Number of fuel spheres in the core (steady state)

452 000 at 60 mm dia

Primary circuit – MPS Pressure Boundary (MPS-PB)

Maximum operating pressure at 100% MCR

9.0 MPa

Design leak rate of MPS 0.1% per day of MPS inventory

Secondary circuit – closed loop cooling

Coolant Demineralized water Closed circuit

Tertiary circuit – Main Heat Sink System (MHSS)

Coolant Sea water

Nominal flow rate 4 000 kg/s

Inlet temperature < 25ºC Shares KNPS Cooling Water (CW) inlet wells

Heat exchange capacity > 230 MW

Outlet temperature < 450 C


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Maximum temperature rise in outlet channel

< 1.50 C Shares KNPS outlet channel

Potable water

Storage capacity 2 270 m3 Fire protection system reservoirs.

Consumption 200 m3 to 250 m3/month Used by demineralization plant, sanitary waste

Staff

During construction Estimated maximum 800

Normal operation Estimated 105

During outage Estimated 250

Operation and maintenance

Plant operating lifetime 40 years

Availability target 95%

General overhauls 30 to 50 days scheduled per 6 years

3.17 Conclusion The proposed Pebble Bed Modular Reactor Demonstration Power Plant (PBMR DPP), which is proposed to be located approximately 400 m to the south of the existing Koeberg Nuclear Power Station (KNPS) on the Farm Duynefontein 34, is a high-temperature, gas cooled reactor technology. It consists of a steel reactor pressure vessel, which contains and supports a metallic core barrel containing pebble fuel spheres. Nuclear energy, which can be regarded as a source of heat energy, is derived from the splitting (fission) of the nuclei of the uranium atoms contained in the pebble fuel spheres. In the case of the PBMR DPP, the heat produced by the fission reaction is transferred to a helium coolant which is then used to drive a turbine that generates electricity. The fuel required for the PBMR will be mined and submitted to a process of uranium extraction and conversion into a feed material for uranium enrichment. After uranium enrichment, nuclear fuel is fabricated, which means that the uranium is built into a structural assembly that is suitable for insertion into the particular type of nuclear reactor under consideration. When the fuel is ‘spent’, the fuel elements are removed and then sent to spent fuel storage. At this point, spent fuel will be sent to a waste repository. This process is known as the nuclear fuel cycle.


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3.17.1 Nuclear Installation License Some of the infrastructure that will be required during the operational phase of the PBMR DPP includes a module building, a generator and main electrical power system, a radioactive waste handling building, a 132kV transmission power line, the extension of the Duine Substation and the widening and construction of some internal and external roads. The transportation of the heavier components of the PBMR DPP is expected to take approximately 3 days and will have significant impacts on traffic flow. Traffic laybye areas (existing picnic areas which will be upgraded) have been identified along the route to allow for the load to move off the road so that traffic can pass. Transportation of fuel to the site has been assessed in a separate EIA process and deemed to pose a low risk as long as environmental management plans and other conditions stipulated in the authorisation are adhered to. In accordance with Section 21 of the National Nuclear Regulator Act 1999, (Act 47 of 1999, “the NNRA”), Eskom has submitted a formal application to the National Nuclear Regulator for a nuclear installation license for the siting, construction, operation, decontamination and decommissioning of the proposed PBMR DPP. Comprehensive technical submissions have and will continue to be submitted to the NNR demonstrating that the proposed plant meets all the NNR’s safety standards. The NNR will review and analyse the technical submissions and then make a recommendation to the NNR Board as to whether or not to issue a nuclear installation licence.

3 project descriptionprojects.gibb.co.za/portals/3/projects/200911 pmbr/j27196... · 2017. 11....

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