From the foregoing, one can derive five principal risk areas for utilities and sponsors of nuclear power projects, i. e., planning, construction, electricity market rates, operational, and waste management/decommissioning (CITI, 2009).

15.4.1 Planning

The lead times from the decision to build a new nuclear power plant until breaking ground are usually measured in multiple years up to a decade. Site selection, acquisition and regulatory site approval, obtaining bids from vendors and bid evaluation, stakeholder involvement and finance arrange­ments are time-consuming steps and often have to be carried out sequen­tially rather than in parallel. Recent early site permits in the US took three to four years from application until the permit was granted. Moreover, nuclear power remains controversial in many jurisdictions and opposition to new developments often results in lengthy hearings and court involve­ments and thus extended planning timelines. Many governments have rec­ognized the added uncertainties in lengthy planning processes and have begun to revise and streamline procedures so as to expedite lead times and reduce uncertainties for nuclear power project sponsors. From a financial perspective, the planning uncertainty faced by developers is the least risky element and no real threat to the financial integrity (CITI, 2009). Still, a utility might have spent several tens of millions on site acquisition, design certification and legal procedures. In addition, a denied site or construction certification after several years may put the utility in a position short of generating capacity with the need to resort to costly alternative suppliers.

The potential impact of nuclear development on local ecology is well docu­mented. The draft NPS sets out some common risks for biodiversity such

as habitat/species loss and fragmentation, disturbance events (noise, light and visual) and air quality concerns. Applicants should develop an environ­mental management plan as part of their EIA. This will be an important aspect of demonstrating to the IPC that mitigation has been adequately factored into the application.

Some practical recommendations for facilitating the preparation of the BIS

are indicated below:

• Develop a solid multidisciplinary team, organised under the direction of a Project Manager, composed of experts who preferably have experi­ence in nuclear projects. If this is not possible in certain areas, they should at least have experience in conventional power plant or large industrial projects.

• Prepare a detailed schedule showing all activities to be carried out and documents to be prepared, issued and reviewed by the different participants.

• Preferably, the BIS preparation team members should work in the same building or close to one another to ensure better integration of the work.

• Prepare the BIS documents using modern information technology (IT) tools in order to facilitate document production, revision and control. It is important to track and control the comments from the different par­ticipants and the changes made in the different versions of the BIS documents.

• The owner should always be in control of the work performed by the external A/E and/or specialised consultant (if any).

• The national nuclear regulatory authorities should be made aware of the work being done to prepare the BIS and be invited to provide their comments to the licensing, nuclear safety and other technical require­ments specified in the BIS.

• It is advisable to hold periodic meetings with prospective bidders to review with them the BIS preparation approach being taken and to get their feedback.

• Existing reactor vendor standard plant designs, when available from the prospective bidders, should be taken into account during preparation of the BIS to ensure that the technical requirements set out in the BIS are realistic and based as much as possible on designs that are actually avail­able on the market.

• It is of the utmost importance to have performed a comprehensive and detailed site study beforehand, so that reliable and complete site data (seismic, geological and environmental conditions, hydrology, cooling water characteristics, grid information, population distribution, social and industrial development) is available. This will enable the bidders to

prepare better bids and subsequently, to proceed more effectively with the plant design phase.

• To facilitate future bid evaluation work by the owner, special attention should be paid in the BIS to giving instructions to the bidders regarding the required structure and contents of their bids. This will ensure that all of the bids have similar structures, contents and information, which makes it easier to review and locate information in the bids.

Each phase in the life of a nuclear power plant requires a license or approval from the RB. Table 20.1, derived from Annex 1 in INSAG-22, describes the main phases and the safety infrastructure needed by countries to establish and maintain a licensing process (INSAG, 2008a). The major phases and the corresponding licensing stages for an NPP are the site, construction, commissioning, operation and decommissioning licenses. Some of these licenses may be divided in sub-stages like ground breaking, first pour of concrete, and the erection of major equipment to facilitate working out the detailed design in parallel with the civil construction work. In that case, the requirements of safety review and submission of technical documents for each sub-stage, together with their submission schedule, should be clearly specified. Conversely, it may be decided to issue the construction and oper­ating licenses in one step, in which case the entire design, including its details, should be submitted before the start of the review process.

Table 20.1 Major activities required by applicants and regulatory bodies at each of the main phases in the life of a nuclear power plant Phase Duration (years) Applicant activities Regulatory activities 1. Design certification 3-4 • Submit documents in accordance with existing regulations and guidance. • Develop requirements and guidance. • Conduct public participation. 2. Siting 2-3 • Select and characterize the site: seismicity, extreme meteorology, flooding, man-made external events, combinations of the above, population density, use of affected land. • Demonstrate compatibility between site and plant. • Submit documents in accordance with requirements. • Develop requirements for siting. • Evaluate compatibility of site parameters with plant design. • Evaluate compatibility of the plant with the site. • Evaluate compatibility of site and plant with emergency planning. 3. Design and construction 5-7 • Develop a preliminary safety analysis report and other required documents. • Construct the plant in accordance with the project and the terms in the license. • Assure quality during design, procurement, assembling and testing. • Develop requirements for design, construction, testing and quality assurance. • Establish a programme of inspections covering equipment fabrication, plant construction, testing and quality assurance. 4. Commissioning 1-2 • Establish an organization to transfer the plant from construction to operation. • Develop a plan for testing and plant acceptance. • Develop the final safety report and apply for operating license. • Prepare operators for accreditation. • Develop requirements for commissioning and operation. • Create a body of experts to witness tests and accept their results. • Establish a process for license reactor operators.

Duration Applicant activities (years)

Regulatory activities

• Operate and maintain the plant in accordance with established requirements.

• Perform periodic testing and inspection on components, systems and structures relevant to safety.

• Conduct self-assessments and external peer reviews.

• Retrofit operating experience.

• Conduct emergency drills.

• Define the need and cause for modifications.

• Analyze the impact of the modifications on the overall safety of the plant.

• Submit the application for approval.

• Conduct the modification in agreement with the applicable standards and conditions.

• Develop documentation for submittal of a request.

• Introduce an ageing management system.

• Review the safety analysis report and the probabilistic safety assessment to prove the validity of the renewal. [109]

• Develop requirements for operation, maintenance, testing, reporting events, conduct self-evaluations, peer reviews and feedback from operating experience.

• Create a continuous oversight system and a body of resident inspectors, enlarged inspection teams and safety evaluators.

• Determine and request the need for plant improvements.

• Be prepared to evaluate plant modifications by establishing guidance for submittals and procedures for evaluation.

• Verify that improvements and modifications comply with the applicable standards and conditions.

• Develop requirements for operation renewal.

• Develop capacity to evaluate the ageing phenomena and their impact on the safety of the plant.

• Develop an inspection and oversight programme to verify the progress of ageing.

• Develop requirements for decommissioning.

• Evaluate the decommissioning plan and the plan to manage radioactive wastes.

• Create a system to inspect and monitor dismantling operations.

Source: derived from Annex 1 in INSAG-22 (INSAG, 2008a).

An operating license is normally issued for the design life of the NPP. However, during the long operation period, which may extend to several decades, the safety status of the NPP is reviewed from time to time, for example by conducting detailed periodic safety reviews. This is to confirm that the NPP, in spite of the ageing of its structures, systems and components (SSCs), meets the current safety requirements and is likely to continue to do so until the next safety review. Towards the end of the license period, if requested by the operating organization, the operating license may be extended for a further period provided a detailed safety review clearly establishes that the NPP can be operated safely for that length of time.

After the NPP is finally shut down at the expiry of the operating license, or due to economic or other reasons, there is likely to be a waiting period to allow for the natural decay of short-lived radionuclides, to reduce the radiation fields on the SSCs to make their dismantling, handling, packaging and transportation to a radioactive waste disposal site easier. Dismantling should never start while fuel is still in the reactor core or in the used fuel decay pool, since as long as nuclear fuel is present an NPP is considered operational and the relevant licensing conditions continue to apply. Even after the fuel is removed from the core and the decay pool, an NPP will have to be kept under surveillance to ensure that there is no undue expo­sure of plant personnel to radiation, and that no unauthorized release of radioactivity is made to the environment. The NPP’s license should be modified appropriately during such periods.

20.10.1 The United States

The example of the United States is relevant as it has been followed, at least partially, by many countries. Moreover, the US regulatory body has maintained cooperation agreements with many other regulatory organizations.

The Atomic Energy Act of 1954, as amended, is the centrepiece of nuclear legislation in the United States. The AEC was an independent agency charged with promoting, licensing and overseeing the peaceful uses of nuclear energy. The Energy Reorganization Act of 1974 abolished the AEC and (in 1975) created the NRC, which was given the authority to grant licenses and provide oversight of safety for nuclear civilian applications.

The NRC maintains two different approaches for licensing nuclear power plants. When the NRC was established, the decision was taken to have a two-step process linking the issuance of a construction permit, followed by an operating license. The licensing requirements under this approach are contained in the Code of Federal Regulations (CFR), 10 CFR Part 50. In 1989 the US decided to adopt a new approach (set out in 10 CFR Part 52, described further on), without abolishing the first.

Any application for a construction permit must be submitted in accord­ance with 10 CFR Part 50. Once received — in the form of a Preliminary Safety Analysis Report — an application is checked for completeness and formally docketed. NRC staff undertake a safety review in accordance with a Standard Review Plan (SRP) leading to a Safety Evaluation Report (SER). The SER is transmitted to a statutory Advisory Committee on Reactor Safeguards (ACRS), which provides independent advice to the NRC on the issuing of a construction permit. Before taking its final decision, the NRC has to conduct, in parallel with the safety evaluation, an environ­mental review of the application and prepare an environmental impact statement (EIS). At the same time, antitrust advice is sought from the US Attorney General’s Office. With all this information, a public hearing is formally conducted and chaired by the Atomic Safety Licensing Board (ASLB), where interested parties may raise questions. Any dissatisfied party can request a review to the US Court of Appeals; otherwise, if the application is successful, the Director of the Office of Nuclear Reactor Regulation issues the construction permit.

The request for an operating license should be requested two to three years before the scheduled construction completion. The Final Safety Analysis Report (FSAR) is the basic document covering this phase. The main purpose of the evaluation is to check that the NPP has been built in accordance with the design approved in the construction permit and that it complies with the applicable requirements. A revision of the EIS is neces­sary, but neither an antitrust report nor a public hearing is conducted, unless formally requested.

The approach described in 10 CFR Part 52 was created to facilitate the standardization of nuclear power plants and simplify the two-step process by unifying the construction permit and the operation license into a single

Construction and Operation License (COL). It also introduced an early site permit and a design certification rule. The early site permit is aimed at resolving site issues, including suitability of the site for emergency prepar­edness and the potential existence of environmentally superior sites.

The design certification recognizes that specific designs comply with established safety regulations. Any applicant for a construction permit or operating license (under 10 CFR Part 50) or a combined license (under 10CFR Part 52) may refer to a certified design and thus ease the licensing process. As in the case with 10 CFR Part 50, the EIS, the antitrust evaluation and the public hearings are maintained.

Personnel will, from time to time, depart from the standards expected of them and adopt behaviours that are not conducive to safe operations and high performance, through incorrect use of tools, failure to use procedures, poor housekeeping, inattention to foreign material hazards, or non-compli­ance with radiological protection protocols, for example. Each time a super­visor or manager observes such behaviours and standards being diminished without correcting them, they have, by their inaction, set a new and lower standard than is desirable.

A common finding in WANO peer reviews and OSART missions is a failure by management to establish and reinforce standards and expecta­tions and a lack of presence of managers and supervisors in the field of work to provide on-going mentorship and coaching of the workforce.

Readers are referred to the WANO guideline GL2002-02, Principles for Excellence in Human Performance.

Analysis of the specialization requirements will be divided into different phases according to the important activities to be accomplished in the nuclear programme from the preparation phase until the commissioning and the initial commercial operation.

Engineering and procurement

To be prepared to issue a bid request for the first nuclear power plant, the staff need to be in place with a basic knowledge of the specific technologies chosen to prepare the bid specification and to establish the evaluation cri­teria. Staff should be available to evaluate and select a winning candidate from a technical, management, business and economic perspective.

Although operators and maintenance technicians do not have to be in place for the moment, some knowledge of operational and maintenance requirements needs to exist within the team. Initial education and training for the remaining resources to fully support plant operation should begin at this time.

The IAEA (2007b) identify the specific human resource needs at this stage including:

• Business and technical expertise for site qualification and preparation of the construction permit request

• Political and social expertise for public communication

• Technical and regulatory expertise to develop and implement regula­tions, codes and standards for plant licensing, site approval, operator licensing, radiation protection, safeguards, physical protection, emer­gency planning, waste management and decommissioning

• Business and technical expertise for fuel cycle procurement and management

• Expertise to conduct training programmes for construction and project management

• Plans to fully staff and train the regulatory body for operational oversight

• Plans to fully staff and train operating, maintenance and support organizations

• Plans to develop future expertise in all relevant areas, including any needed enhancements to the national educational institutions.

Professionals during design periods are needed primarily for project management and engineering. In addition, manpower is required to perform the supporting activities: NPP project planning and coordination, regulatory and licensing activities and fuel cycle activities, among others.

The conceptual design task will involve experienced engineers and tech­nicians. At the end of the conceptual design task all major characteristics of the plant should be defined. The results take the form of systems descrip­tions, conceptual drawings, data compilation and preliminary licensing information. These results should be subjected to an independent review by experienced engineers who are senior professionals not previously involved in the conceptual design development. Consultants who have pre­vious experience on other similar projects may also be utilized.

For basic and detailed design a high level of engineering practice is required. The preparation and review of equipment and component speci­fications constitute an important part of the detailed engineering task. The result of the design work will ultimately be passed on to sub-contractors in the form of equipment and plant specifications and drawings. The produc­tion of these documents is a major effort involving not only the design engineers but also other technical personnel knowledgeable in the areas of manufacturing, materials, engineering, licensing and quality assurance. For specifications work, in particular, there should be engineers with prior expe­rience of writing specifications to lead the task of specifications development.

Table 6.3 summarizes the specialization requirements during the engi­neering and procurement stages.

7.1.2 Plant design and equipment manufacture

A NPP consists of the nuclear steam supply system (NSSS) and the balance of plant (BoP). The NSSS comprises the reactor core and all structures, systems and components (SSCs) required for controlling the reactor power, shutting down the reactor when required and maintaining it in a safe shut­down state. The other SSCs of the NSSS are those that are required for cooling the reactor core during the operating as well as shutdown state and for containment of radioactivity during normal and off-normal operating conditions, including accident conditions. Design, manufacture and con­struction of the SSCs of the NSSS have to meet stringent nuclear standards that require a great deal of specialized expertise and experience. For this reason it is unlikely that the industry in an emerging nuclear power country can undertake this work.

The BoP comprises SSCs that are also found in conventional industry. It may therefore be possible for local design and manufacturing organiza­tions to undertake some of this work. It is, however, to be borne in mind that many of the SSCs of the BoP are also directly or indirectly related to

NPP safety and hence have to be designed, manufactured and tested to high standards. It would be advisable that a careful survey is done in con­sultation with the reactor vendor to establish the feasibility of entrusting specific tasks to the local industry. For some of the tasks the capability of the local industry may have to be suitably augmented. All efforts should, however, be made to maximize the participation of national experts and manufacturing industries with a clear agreement with the reactor vendor. Such participation is very useful for enhancing national capabilities for supporting the future expansion of a nuclear power programme and for undertaking more complex tasks in future, including those related possibly to the NSSS also.

As stated earlier, all NPP equipment and components are to be manu­factured to meet stringent standards. This makes quality assurance (QA) an important part of their manufacture irrespective of the manufacture being done by local or any foreign industry. The NPP owner should there­fore establish early the capabilities and means for ensuring QA during various identified stages of manufacture of all components. The regulatory body should have its own independent capabilities and system in place for carrying out inspection during manufacture for QA. This can be achieved only if a sound infrastructure for QA is available in the country well before any manufacturing activity starts. While some of the QA-related tasks can be outsourced, it is essential that the plant owner as well as the regulatory body have their own technical core groups for reviewing such work.

Enquiries conducted among people living near operating power plants show a positive acceptance of nuclear energy. There are at least three reasons for that acceptance: the socio-economic benefits from the nuclear power plant; growing confidence in the operators through a policy of trans­parent dialogue and information; and the remote perception of a nuclear risk. In this section, the socio-economic benefits of a nuclear power plant are examined.

Most nuclear power plant owners have conducted studies on the socio­economic impacts that they produce in their areas of influence. The US Nuclear Energy Institute (NEI) has so far conducted 13 such studies which include 22 nuclear units, from which some general statements have been published. Other institutions, mainly university departments, have con­ducted analyses for other plants (Exelon, 2008). Moreover, local economic impact assessments have also been conducted for decommissioning (PG&E, 2010) and for the Yucca Mountain nuclear waste repository (University of Nevada, 2003).

The methodology used is based on so-called input-output analysis, avail­able on the market. The application of such models is explained in Section 6 of the NEI analysis of the economic benefits that the institution has so far conducted (NEI, 2008). A number of commercial models are available, with Impact Analysis for Planning (IMPLAN) being the one used in the NEI evaluations.

IMPLAN analyses the interrelations between input-demand and output- supply for any activity, such as the construction, operation and dismantling of a nuclear power plant, within a defined geographical region. The aim of the analysis is to determine the expenditures that the plant will bring to the region, the income generated for local businesses and households, the number of jobs that the plant may provide for the different stages in its life, and the tax revenues generated.

The impacts of the plant on the regional area of interest will include not only direct impacts, i. e. the initial impacts from bringing the plant to the region, but also secondary effects produced by those first impacts, i. e. the demand for goods and services will itself generate new employment and additional spending to deliver the goods and services requested by the plant and by other potential customers. The addition of the two effects is called the total effect, and the ratio between total and direct effects is called the multiplier effect, which can be obtained for each individual effect, such as an increase in jobs, earned income, industry output or revenue from taxa­tion. Multipliers can be obtained for local, county, provincial and state areas.

From the experience obtained through these studies, NEI has published a Fact Sheet which summarizes the contributions of nuclear power plants to state and local economies in the US (NEI, 2010). The values given are normalized averages (normalized to 1 GWe of installed capacity) from the 22 units analysed:

1. Employment. Building a new nuclear power plant will result in the crea­tion of 1400 to 1800 jobs, with peak employment as high as 2400 jobs. During operation each unit generates from 400 to 700 permanent jobs, receiving salaries 36% higher than existing average local salaries. Such an increase in population may generate an equivalent number of local jobs for the goods and services needed.

2. Local economic benefits. The operating plant will require direct goods and services for some $430 million in the local communities and $40 million for labour income, to which indirect effects, amounting to some 7%, have to be added.

3. Federal, state and local taxes. On average, federal tax payments will amount to $75 million, while provincial and local tax revenue will amount to $20 million per year. These taxes are used to create state and local infrastructure.

4. New plant construction. The construction of a new nuclear power plant boosts the supply of commodities such as concrete and steel, and hun­dreds of components and services, such as transportation. It has been estimated that a new nuclear power plant may need some 1 million m3 of concrete and 66,000 tons of steel, 70 km of piping, 480 km of wiring, and 130,000 electrical components. Although the major nuclear compo­nents come from other places, many items could be provided within the area of influence.

Spain has also developed studies on the local socio-economic impacts of nuclear power plants. A study conducted by the Burgos University

Department of Economics at the Nuclenor-owned Garona plant (a 460 MWe GE-BWR) concluded that between 1992 and 2006 the plant spent €80 million on local goods and services, and provided local tax revenues amounting to €6 million. The plant also invested about €16 million per year in plant retrofitting, technological innovation, and research and develop­ment (NUCLENOR, 2007). The study makes an analysis of the social and economic evolution of the population within a 35 km radius of the plant, from before construction to 2007, i. e. after 37 years of operation, by measur­ing unemployment, commercial activities, industrial development, financial entities and the increase of cars and telephones. To separate the impact from the plant, the results have been compared with the corresponding average numbers for the rest of the provincial territory. Factors of 2 have been found for some of the indexes.

Another study was conducted for the Asco and Vandellos plants operated by ANAV (which include three W-PWR, of one GWe each). The study was undertaken by the Rovira i Virgili University (URV) in the city of Tarragona using the output-input methodology described above (ANAV, 2011). It found that output to the Catalonian economy from the activities of ANAV and its workers is four times the initial input during the five-year period from 2004 to 2008. This factor is 3.3 when the territory is limited to the province of Tarragona, where the plants are located. Similar results are obtained for the employment created.

As discussed in Section 9.2, evolutionary reactor designs have concentrated on improving the economics and the performance of existing nuclear reac­tors. Beyond 2030, it is anticipated that new reactor designs will address key issues such as the closure of the fuel cycle or proliferation concerns while possibly ensuring competitive economics, safety and performance.

In order for nuclear energy to remain a long-term option in the world’s energy mix, nuclear power technology development must meet sustainabil­ity goals with regard to natural resource utilization and radioactive waste management. Interest in fast neutron systems and the related fuel cycles has reappeared with the realization of their requisite role in meeting these goals.

Also in alignment with the above sustainability goals, there has been an increased interest in expanding the range of energy products provided by nuclear fission beyond electricity production to include industrial heat, hydrogen and energy for transportation. In this sense, interest in high- temperature gas-cooled reactors that may be able to realize these applica­tions in the most effective manner has also intensified.

Both fast reactors and high-temperature gas-cooled reactors, among others, are innovative reactor designs and the achievement of their full potential is conditional on continued advances in research and technology development.

A few of the topics in which significant work is presently taking place are briefly presented below.

• Core physics: neutronics and thermal-hydraulics. Improvements to safety and economics are met through stringently accurate design require­ments, which must be demonstrated with well-validated calculation tools. Presently, in the area of neutronics, the uncertainties on the nuclear data relevant to several of these innovative reactor designs in many cases are such that they negate the benefits offered by advanced mod­eling and simulation techniques. In the area of thermal-hydraulics, the extreme thermal performance expected from these systems imposes the need for the accurate determination of thermal-hydraulic parameters to relatively high resolution in order to ensure that the relevant safety criteria are met in both normal and off-normal operation.

• Fuels and structural materials. Oxide-, metallic-, carbide — and nitride — based fuels which incorporate depleted, natural and recycled uranium as well as possibly recycled plutonium, and even thorium, are considered for use in innovative reactors. Advanced structural materials with increased strength and creep resistance are sought as a means by which to improve reactor performance by allowing for greater design and safety margins, longer lifetimes and higher operating temperatures; in
parallel, this leads to improved economics through reduction of plant materials and permitted flexibility for design simplifications.

• Coolants and coolant technologies. The coolants under consideration for innovative reactors include alkali metals (e. g. sodium), heavy liquid metals (e. g. lead and lead-bismuth eutectic) and gases (e. g. helium and supercritical water). Each demonstrates distinct advantages and disad­vantages with regard to performance and safety, so much so that serious studies are devoted to all.

Several suppliers and designer organizations are working on the develop­ment of innovative nuclear reactor concepts that address the above con­cerns, although most of them are still in an early stage of development. Some of these are as follows.

• 4S. Toshiba’s 4S (super-safe, small and simple; see Fig. 9.4) is a small sodium-cooled reactor without on-site refueling in which the core has a lifetime of approximately 30 years. Being developed as a distributed energy source for multipurpose applications, the 4S offers two outputs of 30 MWt and 135 MWt, respectively selected from demand analyses. Although 4S has a fast neutron spectrum it is not a breeder reactor since blanket fuel, usually consisting of depleted uranium located around the core to absorb leakage neutrons from the core to achieve breeding of fissile materials, is not provided in its basic design. The reactor power can be controlled by the water/steam system without affecting the core operation directly. The capability of power self-adjustment makes the reactor applicable for a load-follow operation mode. The reactor is a pool type, integral type as all primary components are installed inside the reactor vessel.

Nuclear

• PRISM. The PRISM reactor is a 2200 MWe modular sodium-cooled fast reactor whose development began as a joint project between General Electric (GE) and the US Department of Energy as part of the advanced liquid-metal reactor (ALMR) program. Development has since contin­ued at GE-Hitachi, and the design today incorporates Generation IV objectives. PRISM employs passive safety and uses a modular fabrica­tion technique to expedite plant construction. PRISM uses metallic fuel for better compatibility with the coolant, inherent safety, and ease of fabrication in a hot cell.

Beginning in 2000, 10 countries joined together in the Generation IV International Forum (GIF) to perform the necessary research and develop­ment to support the development and deployment of innovative nuclear energy systems that can be licensed, constructed, and operated in a manner that will provide competitively priced and reliable energy products while satisfactorily addressing nuclear safety, waste, proliferation, and public per­ception concerns (GIF, 2002). GIF experts assessed a large number of candidate reactor concepts by using the common GIF evaluation methodol­ogy resulting in the following six reactor systems selected.

The Gas-Cooled Fast Reactor (GFR, Fig. 9.5) system features a fast — neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference reactor is a 600-MWth/288-MWe, helium-cooled system operating with an outlet tem­perature of 850°C using a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being con­sidered based on pin — or plate-based fuel assemblies or prismatic blocks. The GFR is estimated to be deployable by 2025.

The Lead-Cooled Fast Reactor (LFR, Fig. 9.6) system features a fast — neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The system uses a lead or lead/ bismuth eutectic liquid-metal cooled reactor. Several options with different plant sizes have been proposed including a battery of 50-150 MWe that features a very long refueling interval, a modular system rated at 300-400 MWe, and a large monolithic plant option at 1200 MWe. The fuel is metal or nitride-based, containing fertile uranium and transuranics. The LFR system is estimated to be deployable by 2025.

The Molten Salt Reactor (MSR, Fig. 9.7) system features an epithermal to thermal neutron spectrum and a closed fuel cycle tailored to the efficient utilization of plutonium and minor actinides. In the MSR system, the fuel is a circulating liquid mixture of sodium, zirconium, and uranium fluorides.

The molten salt fuel flows through graphite core channels, producing a thermal spectrum. The heat generated in the molten salt is transferred to a secondary coolant system through an intermediate heat exchanger, and then through another heat exchanger to the power conversion system. There is no need for fuel fabrication. The reference plant has a power level of 1000 MWe. The system operates at low pressure (<0.5 MPa) and has a coolant outlet temperature above 700°C, affording improved thermal effi­ciency. The MSR is estimated to be deployable by 2025.

The Sodium-Cooled Fast Reactor (SFR, Fig. 9.8) system features a fast — neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. Two options have been envisioned. The first is an intermediate-size (150 to 500 MWe) sodium-cooled reactor with a uranium-plutonium-minor-actinide-zirconium metal alloy fuel,

supported by a fuel cycle based on pyrometallurgical processing in colo­cated facilities. The second option is a medium to large (500 to 1500 MWe) sodium-cooled fast reactor with mixed uranium-plutonium oxide fuel, sup­ported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. The outlet temperature is approxi­mately 550°C for both. The SFR is estimated to be deployable by 2015.

The Supercritical-Water-Cooled Reactor (SCWR, Fig. 9.9) system features two fuel cycle options: the first is an open cycle with a thermal neutron spectrum reactor; the second is a closed cycle with a fast-neutron spectrum reactor and full actinide recycle. Both options use a high-temperature, high — pressure, water-cooled reactor that operates above the thermodynamic critical point of water (22.1 MPa, 374°C) to achieve a thermal efficiency approaching 44%. In either option, the reference plant has a 1700-MWe power level, an operating pressure of 25 MPa, and a reactor outlet tempera­ture of 550°C. Passive safety features similar to those of the simplified boiling water reactor are incorporated. The balance-of-plant is considerably

llll

Generator

Purified salt

Pre I cooler

Heat sink

Intercooler

Compressor

Emergency dump tanks

9.7
GIF Molten Salt Reactor (MSR) generic concept (illustration courtesy of Idaho National Laboratory).

9.8 GIF Sodium-cooled Fast Reactor (SFR) generic concept (illustration courtesy of Idaho National Laboratory).

Electrical power

;o Pump

9.9 GIF Supercritical Water Cooled Reactor (SCWR) generic concept (illustration courtesy of Idaho National Laboratory).

simplified because the coolant does not change phase in the reactor. The SCWR system is estimated to be deployable by 2025.

The Very-High-Temperature Reactor (VHTR, Fig. 9.10) system uses a thermal neutron spectrum and a once-through uranium cycle. The VHTR system is primarily aimed at relatively faster deployment of a system for high-temperature process heat applications with superior efficiency (see Section 9.5). The reference reactor concept has a 600-MWth helium-cooled core based on either the prismatic block fuel of the Gas Turbine-Modular Helium Reactor (GT-MHR) or the pebble fuel of the Pebble Bed Modular Reactor (PBMR). The primary circuit is connected to a steam reformer/ steam generator to deliver process heat. The VHTR system has coolant outlet temperatures above 1000°C. The system may incorporate electricity generation equipment to meet cogeneration needs. The system also has the flexibility to adopt U/Pu fuel cycles and to offer enhanced waste minimiza­tion. The VHTR system is the nearest-term hydrogen production system, estimated to be deployable by 2020.

. Heat Reactor He|ium exchanger coolant