EPRI, in collaboration with USDOE, have developed a set of requirements to establish the technical basis for the design of advanced light water reactors (ALWRs) (IAEA — TECDOC-968, 1997). A first objective is to establish a basis for licensing future LWRs, including the resolution of outstanding severe accident issues, and to gain agreement with the USNRC. Secondly, there is an objective to provide a standardised plant design with vendor certification. Thirdly, there is an intention to provide a set of technical requirements, to minimise the risks to investors in completing and operating the first ALWR.

The EPRI Utility Requirement Document (URD) covers top-level programme policy statements and detailed requirements for specific ALWR designs. It includes large evolutionary systems with improved active safety systems and also passive system designs including natural circulation, gravity-driven refill and stored energy as essential safety functions. Both passive PWR and BWR systems are included.

The document was first published in 1990 and has been used in the development of several new LWR designs. It has been developed by the US utilities to reflect the procedures’ rules, regulations, codes and standards of the US. However, there have also been contributions from European utilities, which have developed their own set of standards, as discussed below.

A group of 10 countries (Argentia, Brazil, Canada, France, Japan, Republic of Korea, South Africa, Switzerland, UK and US) are working together specifically to develop a roadmap to pursue R & D on future Generation IV systems (Generation IV Nuclear Energy Systems, 2003). These innovative reactor systems are described in subsequent chapters. The GIF initiative is relatively recent, starting in January 2000 and initiated by the United States Department of Energy (USDOE). The objectives are to develop future reactor systems that are competitively priced, while addressing safety, waste, proliferation and public perception concerns. The reactors cover water, gas-cooled thermal and fast spectra, liquid metal (sodium, lead and lead-bismuth) cooled and molten salt designs.

The issue of in-vessel FCIs has been postulated in the context of present generation reactors. The defence strategies are an attempt to exclude this possibility by design or to demonstrate that the vessel will not fail or demonstrate that the containment remains intact after vessel failure. In advanced reactors, if failure of the vessel is assumed, there is the opportunity to design a reactor cavity that can survive the load (EIBL et al., 1992), and also to protect the containment from flying missiles by including an upper shield or slab.

It is generally expected that there may be a greater possibility of vessel failure if the system has been depressurised. Depressurisation is often a strategy in plants with passive injection to insure that injection can occur and so in principle in-vessel FCIs may be an issue for some advanced plant designs. However, some analysts believe that steam explosions in-vessel will not be sufficiently energetic to cause vessel failure.

The possibility of ex-vessel FCIs can be substantially reduced by preventing molten core material exuding from the bottom of the vessel from coming into contact with water. A number of preventative features are proposed in current advanced designs.

The ‘core-catcher’ has been proposed for the EPR, for example, Figure 11.3. In this design, the melt is spread horizontally over a large dry area of about 150 m2. Once in this

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Figure 11.3. European pressurised reactor. Source: Leverenz (1999).

spreading compartment, the corium would then melt through various low melting point plugs that would eventually let water through from a large IRWST tank to flood the corium (Leverenz, 1999). Heat would be dissipated from the melt by evaporation for a 0.5-1 day period, after which an alternative containment cooling system would come into operation. In the case of EPR, this involves containment sprays, cooling of the water in the spreading compartment and also cooling of the IRWST water.

Other types of core catcher have been proposed. These include a similar ‘plug melt — through’ concept into crucibles in a dry vertical core catcher concept. The crucibles are then cooled by natural circulation of water, which is ultimately discharged through the containment to an ultimate heat sink. Another type radiates heat to a large conducting surface in the reactor cavity, which is then cooled by external natural circulation.

The retention of core melt has been investigated in several experimental programmes. This includes programmes in Germany and the MACE tests in the US.

In another German design (Kuczera, 1992), the corium is allowed to fall into a dry cavity with a thin bottom layer of low melting point material. Hollow plugs are eventually uncovered allowing water to flow up the plug holes and cover the corium.

Another variation of design to achieve cooling is to have a staggered pan arrangement in an oxidic ceramic bed. The upper part of this bed remains dry and the lower part is flooded with water. Heat is extracted via natural circulation of water through the particle bed. The possibility of steam explosions is reduced because the top part of the bed remains dry.

The other way to ensure that melt does not come into contact with the water is to prevent the vessel failing. One postulated approach is to flood the vessel in the reactor cavity. The effectiveness of this measure will depend on the power density and the geometry of the vessel (surface area). In this method, heat is removed from the melt via conduction through the lower head of the vessel.

Different fuel and fuel cycle concepts have been considered in the reactor, arranged in a sub-critical state. Some of the primary areas of research at various laboratories are shown in Table 13.2. These are expanded further in Section 13.8.

Brookhaven National Laboratory (BNL) has focused mainly on fast spectrum concepts, liquid sodium cooling and oxide or metal solid fuels based on sodium-cooled fast reactor technology. Also in that laboratory, particle bed/bead fuel has been investigated in the thermal spectrum, as considered in space propulsion reactor technology.

In Japan, the Japanese Atomic Energy Research Institute (JAERI) has concentrated on MA burning in a fast neutron spectrum, with solid fuel also based on sodium-cooled fast reactor technology, or molten chloride fuel, as yet an unproven technology.

Los Alamos National Laboratory (LANL) has developed a concept based on a thermal neutron spectrum with molten fluoride fuels with different fissile materials such as weapons grade plutonium, LWR spent fuel (minor actinide and fission products) and thorium fuels based on molten salt water reactor technology. Liquid lead-bismuth systems in the fast spectrum have also been considered.

The CERN group in Geneva, Switzerland, has put forward the concept of solid ThO2/233UO2 fuel in a fast spectrum based on liquid lead/liquid lead-bismuth reactor technology. This uses a cyclotron-based system. The applications are for energy production or waste transmutation.

At the Institute of Theoretical and Experimental Physics (ITEP) in Russia, different technologies for the conversion of weapons plutonium and long-lived radioactive waste are being considered. These include heavy water suspensions, molten fluoride and liquid lead fast spectrum systems.

Work is also being carried out in various laboratories within the EU, including France, Germany, Italy, Sweden and the UK. Within France for example, research, carried out at CEA has focused on options for radioactive waste management.

There are clearly many options under investigation in the international community, which offer a reprocessing capability for nuclear fuel and in the case of weapons plutonium, a means for the reduction in the world’s stockpile of plutonium. Further assessment of the various options is continuing.

The utilisation of lead-bismuth reactors for district heating or for seawater desalination is being investigated in Russia (IAEA-TECDOC-1056, 1998). There are 150 reactor-years of experience of lead-bismuth reactor technology in Russia from application in the country’s submarine fleet.

The technology is now being reassessed for either co-generating or single application district heating or desalination plants. Some of the plants, e. g. ANSTREM could also be used for refrigeration applications. The coolant has desirable chemical, activation and thermophysical properties, including low chemical reactivity with water, low long-lived induced gamma activity, negative void coefficient, a high boiling point and low freezing point. Some possible new designs for low temperature applications are shown in Table 14.7. The technologies under consideration include modular and small transportable plants such as ANSTREM with a compact reactor layout, but also larger plants such as BREST 300.

15.7. PASSIVE HEAT REMOVAL SYSTEMS

There are some phenomena that occur in evolutionary reactors under accident conditions that assume greater significance compared with presently operating plant. These mainly

relate to passive system performance, including natural circulation and passive injection, and also decay heat removal from large water pools. A major experimental programme to investigate these phenomena was carried out by Westinghouse leading up to the design certification of AP600 (Squarer et al., 1988; Venne et al., 1992) to confirm the conceptual design. Other recent programmes are shown in Table 15.6.

Space reactor systems have been studied since the early days of nuclear power in the late 1950s. However, only one US reactor (SNAP-10A) (Harman and Susnir, 1964) and a few Russian reactors have ever been in space. There is now some renewed interest in nuclear power for space missions, in the US and also Europe.

In general, for space applications, fast reactor gas or liquid metal cooled designs, operating at high temperature are the most appropriate to meet the various requirements and in particular, launch constraints. Clearly also reliability is important and this depends on the status of the possible technologies.

Space applications include, planetary base applications, e. g. for Mars or the moon, nuclear propulsion and radioisotope power systems (RPS). For the former, possible designs include the lithium liquid metal cooled concepts, SP-100 in the US (Sapir et al., 1987) and the ERATO system in France (Carre et al., 1987), these generating power in the range 100-500 kWe. Gas cooled systems include the Sandia National Laboratories Dual Purpose design (Lipinski et al., 1999), and a Russian Project 1172 gas-cooled design (Andreev et al., 2000). A low-power PWR water-cooled system has also been investigated by Technicatome.

For propulsion, many of the reactor concepts under consideration have been developed from other applications. In general many reactor systems that have been developed to supply electrical power, can be employed as a power source in a nuclear electric propulsion (NEP) systems. The SP 100 and ERATO system could be adapted. The UK 200-SNPS was a particle bed system, designed for earth orbit electrical power supply, but could be adapted. The Enabler NERVA (Livingston and Pierce, 1991) was primarily aimed at nuclear thermal propulsion (NTP), where the energy source heats the propellant directly (as opposed to NEP where electrical power from the reactor is used for accelerating the propellant). The Russian TOPAZ-2 liquid metal (NaK) cooled system (Voss et al., 1991) or more advanced TOPAZ concepts could be used. There are also combined cycle (NEP&NTP) nuclear propulsion and other advanced concepts under consideration.

Some of the reactor designs are such that the same generic design can be used for both planetary base and propulsion applications. An example of one such is the ESCORT Derivative reactor (Feller and Joyner, 1999), designed for in-space propulsion and power (25 kWe) and to supply 160 kWe for 10 years on the surface of Mars.

Finally RPS consisting of a nuclear radioisotope heat source and power conversion, have been developed. This technology started in the SNAP programme in the 1950s and culminated in the General Purpose Heat Source (Angelo and Buden, 1985) module flown on the Galileo and Ulysses spacecraft. RPSs typically generate a few kilowatts.

Space nuclear reactor programmes are being supported by the National Aeronautics and Space Administration (NASA) (Nuclear Reactors in Space) and the European Space Agency (ESA) programme. A review of space nuclear power and propulsion for future space exploration is given in (Bond and Sweet, 2003). A particular interest at present is the benefits of nuclear power systems for Mars exploration (Sweet et al., 2002). In particular, work is on-going to examine the feasibility of different reactor systems, including the feasibility of a small gas-cooled, particle bed reactor, to power a Mars mission.

1.4.1 Magnox Reactors

The first Magnox reactors built in the UK were at Calder Hall and Chapelcross. These were just 50 MW plants; eight units being built in total which were commissioned between 1956 and 1960. These first plants were originally envisaged for the purpose of producing plutonium but were also operated to produce electricity. They were followed by a series of higher rated plants commissioned between 1962 and 1971. The most highly rated plant was Wylfa at 590 MW operating at a gas pressure of about 27 bars. Many of the earlier plants are now shut down but the later plants are still in operation.

The Magnox reactor core consists of a ‘pile’ of graphite blocks or bricks which contain channels. Carbon dioxide at a pressure of typically a few tens of bars flows through these channels, which also contain the fuel elements or control rods. The fuel elements consist of natural uranium bars clad with a magnesium alloy known as Magnox. These are machined into a ‘herringbone’ pattern in order to optimise heat transfer. A metallic fuel was adopted; i. e. natural uranium was used. The magnesium alloy was specifically chosen because it did not have a significant absorption of neutrons, enabling natural, rather than enriched uranium to be used.

Typical geometric and operating parameters are defined to limit the internal temperature of the elements to about 650°C, a critical temperature at which deformation of the uranium crystal lattice occurs. Similarly, the can temperature is limited to 420°C, associated with the use of Magnox alloy. A typical Magnox core is about 8 m high and 14 m in diameter. The core exit gas temperature is about 400°C.

On exiting the core, the coolant flows directly to the steam generator and then is pumped back to the reactor. The efficiency of the steam cycle is around 31%.

In the early Magnox designs, the vessel was made of steel and the steam generators (heat exchangers) were external to the pressure vessel. In Oldbury and Wylfa, the heat exchangers were placed inside the pressure vessel, constructed with pre-stressed concrete (Smitton, 2000).

Magnox reactors have in general operated very successfully in the UK over a period of many decades. However, from an economic perspective they have a low power density with high fuel costs.

The accident at TMI-2 set in motion action plans for improved operator training in many countries (U. S. Regulatory Commission, 1980). A particular consequence was to recognise that training must be broadened to include more emphasis on operational incidents and accident circumstances. Further training should not just be for operators but training of maintenance and all personnel connected with the plant operation should also be included. Recommended practices are given in Table 3.4.

Training needs to be provided at various levels. It must include new staff but also employees who might be changing jobs through transfer, redeployment or promotion (Leclercq, 1986). It should occupy a reasonable amount of time; e. g. French nuclear power plant workers receive typically 80-90 h training per year.

Present day training includes extensive use of simulators and electronic-aided teaching procedures. These are used to teach the candidates the single component aspects of plant operation, e. g. the chemical and volume control system, control of the reactor and managing the balance of plant function (e. g. the turbine-generator performance).

Full scope simulators provide an exact replica of the control room and enable a complete simulation of plant behaviour. The candidate can, therefore, be subjected to all forms of plant condition from start-up, shutdown and response to faults.

The training of maintenance staff is equally as important as the training of operators. Their training involves acquisition of the necessary technical expertise but also they need to learn the organisational and communication skills to manage a large workforce. They must acquire the necessary skills to work and be compliant with the various safety instructions and other working conditions, including working under the pressure of meeting challenging delivery deadlines.

The training of power plant personnel has been addressed by the IAEA. A guidebook was published which was concerned with the training to establish and maintain qualified and competent operations’ personnel for plant operation. The guidebook has been subsequently updated in 1996 (IAEA, 1996) and the latest version includes worldwide experience that has been gained since earlier publication.

IAEA recommendations are that there should be a systematic approach to training for nuclear power plant personnel. The approach should not only cover the operators but should cover the role and responsibilities of management including the training of management. The approach should include the evaluation of training methods and look for ways of making training more effective. It should also cover the organisations involved in providing the training.

Significant areas of research to support the present generation programme include safety and performance at all stages of the fuel cycle, reactor safety during plant operation, radioactive waste management, radiological protection, and other activities to benefit from ‘lessons learned’ in the past. There are active work programmes in all these areas. For example within the European Union there have been numerous activities funded by the European Commission Euratom Programme and corresponding counterpart national programmes (European Commission, 1994).

Regarding the fuel cycle, a key objective of the European Programme is to explore innovative approaches. Alternative fuel cycle concepts are being considered, primarily to address the problems of safeguarding long-lived radioisotopes. A particular example is partitioning and transmutation (P&T) which aims to provide a process of reducing the level of long-lived radioisotopes in high level waste. These methods rely on complex

separation techniques and methods of transmutation that have yet to be developed. These techniques are in an early stage of development and are not yet prototyped at the industrial scale. The research programme is considered later in the book.