15.1. INTRODUCTION/OBJECTIVES

In connection with design optimisation, comprehensive experimental and theoretical programmes for component and integral design, testing and certification are carried out by vendors. In addition, safety-related projects are performed by major research institutes within various national and international programmes. Many data available for present generation plants are still applicable to evolutionary and advanced future reactors. However, additional data are required for certain physical processes which occupy increased significance in advanced passive systems, e. g. natural circulation, condensation and non-condensable gas-related phenomena. Relevant experimental research for both current and future reactor systems is summarised for some of the major designs under review in this book. In regard to international activities, particular reference is made to the extensive EC research programmes (FISA 2001, 2001; FISA 2003, to be published). The focus in this chapter will be on experimental programmes; theoretical work is covered in the next chapter.

The majority of research to date has been in support of the present generation and evolutionary plant. Many present day plants are coming up towards the end of their original design lives and with the dearth in new build there are strong incentives to consider extension of life. This has, therefore emerged as a key area of research. Severe accident research is being used to develop severe accident plans and support Level 2 PSAs; hence severe accident research is also prominent. Research specific to evolutionary passive designs includes work on natural convection cooling, with and without the presence of water. Many of the evolutionary designs also include improved provision against severe accident loads; research is carried out on passive heat removal from melts within the reactor cavity.

Significant areas of research to support the present generation programme include safety 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 EC Euratom Programme and corresponding counterpart national programmes.

In addition to research projects per se, there are a number of joint and other collaborative projects being co-ordinated under the auspices of the NEA, which primarily collect, make data available and perform analysis on the data. Some of these projects are of a general nature, e. g. the International Common-cause Data Exchange (ICDE) project

collects operating data related to common-cause failures. The Fire Project collects data related to fire events in nuclear environments and the OECD Piping Failure Data Exchange (OPDE) project collects and analyses pipe failure event data. More information is given in NEA Annual Report (2002).

There are a number of technical issues that will need to be addressed in regard to the innovative reactor systems that are envisaged for the future, e. g. the Generation IV concepts. These are covered briefly at the end of this chapter. Many of the concepts will require significant research effort. For some systems, R&D activities are already underway, e. g. in regard to the supercritical and high-temperature gas systems that are expected to be among the first Generation IV systems to become available.

In the first part of this chapter, research primarily relevant to current generation plant is considered.

Reactor kinetics codes are used to calculate assembly averaged neutron flux and power distributions in a reactor core under transient conditions. The UK PANTHER code developed by British Energy is a typical example (Hutt, 1996). It includes a neutron diffusion neutronics model, coupled with a 1D thermal hydraulics model for the core region. The code has also been coupled with the RELAP5 system thermal hydraulics code, see below, to provide a neutronic/primary circuit modelling capability. The code can perform reactor calculations, fuel management studies and safety transient analysis. It can also be used for on-line calculation support. Other codes include: RAMONA, PARCS (Joo, 1998), SIMULATE-K, CORETRAN and SAPHYR.

Whole-core events, such as macroscopic temperature changes cause global power changes and these can be modelled adequately with point-kinetics models. These models require as input, reactivity coefficients, the effective delayed neutron fraction, generation time and control rod worths. However, in some transient conditions such as rod ejection or control rod drop, localised events occur that require multi-dimensional neutron kinetics analysis with codes of the type mentioned above. One-, two — and three-dimensional models require neutronic input parameters such as assembly averaged neutron cross­sections and delayed neutron fractions that are obtained from the nuclear data codes. The neutronics codes typically model energy groups condensed into two energies. Delayed neutron fractions would usually be modelled on a nodal basis.

A review of applicable existing thermal gas cooled reactor experience and previous gas cooled fast reactor projects is given by Mitchell et al. (2001). Gas reactor physics methodologies have been established, e. g. WIMS and PANTHER for application in the UK gas reactor industry. Gas reactor physics methodologies are being extended to high — temperature reactor applications with pebble bed fuels, taking advantage of already existing experience. The fuel and core design for gas cooled fast reactors are necessarily different, e. g. the graphite pebble bed concept cannot be used because graphite is a moderator and also because of the fast neutron core reactivity sensitivity to geometry variation. The fast gas reactor core will probably be based on more conventional LMFBR design using MOX or UOX steel clad pellets, e. g. as in the ETGBR design.

A range of computer codes has been developed for fast reactor neutronics (IAEA-TECDOC-1083, 1999). These include codes based on classical diffusion theory, transport theory and Monte Carlo methods. Within the European fast reactor community, the European reactor analysis optimised system (ERANOS) code system has been developed. This system embodies a modular system of codes not only for performing neutronic system design calculations but also for experimental analysis of critical facilities.

Undoubtedly, more research will be needed to develop reactor physics methodologies for the evolutionary plants and certainly for the more innovative concepts. However, there already exists substantial pool of experience on which to build.

The first such reactor to generate electricity was the US Experimental Breeder Reactor 1 (EBR 1). This started in 1951 with a capacity of 200 kWe. It was fuelled by highly enriched uranium-235. In common with future fast reactor designs, the core was small and compact. The fuel pins were just 1.25 cm in diameter. The core consisted of 217 pins in a hexagonal lattice. The coolant was a sodium/potassium alloy, surrounding the central region was a blanket region containing rods of natural uranium. EBR 1 operated until 1963 and yielded considerable information on liquid metal fast breeder reactor (LMFBR) technology. A second reactor EBR 2, 15.7 MW, was also built on the Arco site in Idaho.

A 60 MW commercial reactor, Enrico Fermi 1 went critical in 1963. This reactor underwent a serious loss of coolant accident in 1966. It restarted for a few years but was finally shut down in 1970.

The US fast reactor programme continued with various test facilities until 1983, e. g. the southwest experimental fast oxide reactor (SEFOR) at Arkansas, the transient reactor test experiment (TREAT) at Argonne and the fast flux test facility (FFTF) at Hanford.

Within Europe, the United Kingdom atomic energy authority (UKAEA) built several research reactors before the Dounreay fast reactor (DFR) was commissioned and became critical in 1959. DFR had a modest electrical capacity of 14 MWe. It was closed down in 1977. The prototype fast reactor (PFR) had an electrical output of 254 MWe and entered service in 1975. It operated for over a decade before being shut down.

This sodium-cooled fast reactor was a pool type design. A pool of sodium is contained in a vessel with sodium pumped through the core by pumps contained within the pool. The hot sodium then passes through an intermediate heat exchanger; transferring heat to a second sodium-cooled loop. The latter transfers heat to a water/steam loop via the steam generator. This tertiary loop system ensures that any radionuclides produced in the primary vessel remain in the vessel and are not transferred to the steam generator.

In this type of reactor design, the reactor functions on fast neutrons, there is no moderator.

In France, a similar 250 MW prototype was also built (Phenix), which was then followed by a commercial sized plant (Superphenix), the latter commissioned in 1986 (but now closed down permanently).

Other countries have explored the production of fast reactors, e. g. Germany, Japan, India and the former Soviet Union.

LMFBRs have a number of advantages. Liquid metals have desirable thermophysical properties. The coolant has a low melting point, coolants can be chosen, e. g. sodium and potassium, which have low neutron absorption. Sodium has a high thermal conductivity, albeit a lower specific heat than water and it has a high boiling point, etc.

LMFBRs also suffer from a number of disadvantages and problems. There are concerns over the use of sodium since it is highly reactive to oxygen and water. There is a potential problem of isolation of the sodium and water-cooling loops. There have been problems in the steam generators of fast reactors.

In recent years the development of fast reactors at the commercial scale has slowed down. Nevertheless, the potential for fast reactors exists and is still under review in some countries. Fast reactors are again under consideration in the US Generation IV programme.

Historically, the fast reactor has always been considered in relation to its fuel cycle, its ability to burn and breed plutonium. In addition, most reactors produce plutonium, in differing amounts, which can in principle be recovered for utilisation in a fast reactor fuel cycle. However, there are safety and economic issues associated with fuel reprocessing, these are considered later. Plutonium can also be burnt in thermal reactors to improve the economics of the thermal fuel cycle.

A commonly accepted principle is that the direct responsibility for safety of a nuclear plant rests with the utility. This contrasts the role of the regulator whose function is to set the safety goals and to ensure these are met. As for differences in regulatory focus across different countries, there are also differences in relationship between utility and regulator. This relationship is more formal in some countries than others. However, in most countries, there is a desire from both utility and regulator to promote an increasingly collaborative working relationship and encourage open dialogue between the two parties.

It is the responsibility of the licensee to have in place emergency operating procedures for the plant and also emergency planning procedures for the whole nuclear plant complex (Pershagen, 1989). These include instructions for plant operation for accidents within the design basis but also procedures for severe accidents beyond the design basis (Table 3.7).

Operating rules for design basis accidents (event oriented)

Emergency operating procedures for beyond design basis severe accidents (symptom oriented)

Pershagen (1989).

Emergency planning procedures include the establishment of an organisation to implement the plan, including any required accident management actions. Emergency planning procedures beyond the plant boundary usually are the responsibility of other local and/or national authorities. However, the licensee must be prepared to liase and co-ordinate operations with these authorities.

The utilities are regularly audited by safety authorities to ensure that all the required frameworks/organisations are in place for the continued safe operation of their nuclear plants. An important principle is that the safety case is ‘owned’ by the plant, i. e. that a utility has in place a sufficient number of adequately trained staff who understand the relevant issues and are suitably qualified and experienced personnel (SQEP).

Reactor vendors clearly perform an integral part in ensuring the operational safety of a plant. They may be called upon by the utility to provide services not just at the design and initial licensing stages but also during the lifetime of the plant. There may be a requirement to back-fit more efficient safety systems, or to purchase fuel from a new fuel vendor (the latter is more performance — than safety-related). In the US, there has been a push by some vendors to license specific designs with the USNRC. It is expected that this would substantially simplify the licensing process of that particular design, e. g. in countries outside the US.

Waste management issues are discussed in more detail in the next chapter. Some countries have already put in place schemes for the disposal of high level waste in geological repositories; others have not yet committed to this approach.

There are additional safety issues associated with the storage of spent MOX fuel since MOX fuel generates more heat than UO2. It may, therefore, be necessary to down-rate dry waste storage. A further point is that storage pools may require additional neutron poison to ensure adequate sub-criticality.

8.1. INTRODUCTION/OBJECTIVES

Advanced reactors will need to meet continued demands for increased safety. This chapter reviews present legislation and possible future licensing requirements for the safety of advanced future reactor operation. The current generation of nuclear plants was designed to withstand accidents from a set of ‘design basis’ events. Most countries set limiting core damage frequencies and limiting probabilities for large fission product releases. An objective of many designers for advanced plants is to extend the current design basis to include accidents of increased severity and lower probability to meet expected more stringent future regulatory safety requirements.

The main focus of this chapter will be on the licensing and safety requirements for evolutionary reactors. Many regulators believe that the national frameworks already in place for existing plant remain adequate for evolutionary plant. However, there are increasing endeavours by international bodies such as the EC and IAEA to promote more harmonised agreement on nuclear safety criteria and therefore encourage a more harmo­nised approach to licensing in their member states. There is similar encouragement from the industry side with the development of standardised utility requirements (URs) for member states, e. g. the US and European URs described in the previous chapter.

At the present time, nuclear power supplies only a small contribution to the country’s energy requirements, generating 2.9% of the total (World Nuclear Association, 2003). Currently operating there are an old PHWR (125 MWe) supplied by Canada and the Chasnupp PWR (300 MWe) supplied by China. Both these reactors are operating under international safeguards. Pakistan also has a 9 MW research reactor.

There are also plans for another Chinese-designed PWR. This is proposed as a second unit at the Chasnupp nuclear plant at Chashma (Foratom e-Bulletin, 2003b).

The technologies that have been investigated for a gas-cooled fast reactor concept have been reviewed by Mitchell et al. (2002). The basic idea has been to extend the thermal reactor concept but with a fast reactor core. The existing technology gas-cooled breeder reactor (ETGBR) was based on an AGR technology with a carbon dioxide-cooled system. The gas-cooled fast reactor (GCFR) design of General Atomics takes a helium technology as its basis. The gas breeder reactor (GBR) covered four different design concepts in respect of carbon dioxide and helium as possible coolants, oxide pins vs. particle fuel, etc., see below. These are surveyed in Table 12.4.

12.6.1 LFR (Gen IV)

Lead and lead-bismuth systems are being considered in the GIF programme (The US Generation IV Implementation Strategy, 2003; Figure 12.5). Examples are listed in Table 12.6.

The system is based on natural convection cooling with outlet temperature 550°C. It could be somewhat higher ~800°C subject to improved materials development. It can be used within a long life closed fuel cycle of up to 30 years in some concepts. It is anticipated to be used for electricity production, hydrogen production and actinide management.

In these processes, low-temperature steam is taken from the power plant turbine of the supplying plant to heat the saline solution. In commercial distillation, there are a number of heat recovery stages in series, because of the high heat of evaporation of water. These stages are at progressively lower pressures, resulting in flashing and mechanical vapour compression to occur.

In general, the more stages in place, the more efficient is the process. The number of stages is limited by both economic and technical reasons, e. g. the overall temperature

difference between the heat source and the cooling water sink. The typical temperature reduction per stage for a commercial plant is 2-5°C. In terms of thermodynamic efficiency, expressed as kg of water produced against kg of steam used, the figure is 6-10 for MSF applications and up to 20 for MSD. These processes are described below.

14.4.1.1 MSF Distillation. In this process, seawater is passed through a number of stages where it is progressively heated (see below) until it reaches the main heating section supplied by the process heat source, see for example (IAEA-TECDOC-1056, 1998). The brine is then returned through these stages and freshwater is eventually obtained through a series of flashing and condensation processes. In particular, as the heated brine returning from the heat source passes into the first stage heat recovery section, flashing occurs due to pressure reduction. Vapour is produced which condenses on the entry pipe-work to the heating section within the first stage (providing the progressive heating referred to above). The condensate is collected in trays. This condensate together with the remaining brine (that has not flashed) is passed on the second stage. The process is then repeated for a number of stages and the separation process completed. Non-condensable gases are removed by a steam-jet ejector system. The seawater is also chemically treated to remove scale.

14.4.1.2 MED. This process also consists of a number of heat-exchange sections. At the first stage steam from the heating boiler passes through a tube bundle which is cooled by evaporating the entry seawater on the other side of the tube bundle. The resulting steam is then passed to a second stage heat exchanger. Any seawater not evaporated at the first stage is passed on to the second stage. The process is then repeated to complete the separation process. MED plants require similar scale removing processes as do MSF plants.

Several designs have been used. The main difference is in the design of the heat exchangers. The low-temperature horizontal tube multi-effect process (LT-HTME) has horizontal tubes and the brine is sprayed over the outside of the tubes. In the vertical-tube evaporation process (VTE), the evaporation is inside vertical tubes. The LT-HTME is the more dominant process used.

In general, MED plants are more efficient than MSF plants because their heat transfer processes are more efficient for given heat transfer area and similar temperature difference between the heat source and cooling water.

14.4.1.3 RO. RO is also used as a separation process (IAEA-TECDOC-1056, 1998). This process has been applied commercially and can produce freshwater down to between 100 and 200 ppm of total dissolved solids. The electricity consumption is in the range 4-7 kWe h m_3.

In this process, seawater (brine) and water are held in a vessel in two-solution compartments separated by a semi-permeable membrane. Pressure is applied to the compartment containing the brine, sufficient to overcome the natural osmotic pressure of the solution and the permeate pressure (NB this is negligible compared with the natural osmotic pressure). In these circumstances, water flows from the brine compartment to the water compartment, the brine become more concentrated and purified water is obtained in the water compartment.

As the seawater is fed into the brine compartment, it is compressed up to 70-80 bars, sufficient to overcome a natural osmosis pressure of the saline solution of about 60 bars. In practice, only a portion of this water flows through the membrane, the remainder is discharged. The flow through the membrane is proportional to the pressure gradient of the applied pressure less the solution osmotic pressure. The proportionality factor depends on a range of factors including the geometry (shape, area, thickness) and the chemical properties of the membrane, the pressure, concentration, pH and temperature. Membranes have been used of varying design, spiral-wound, hollow fibre, also tubular, plate and frame type, the former two designs being the most commonly used.

14.4.1.4 Hybrid Desalination. Hybrid desalination systems can be used to combine power generation, with MSF or MED, and RO processes. This combined capability can be utilised to advantage in different ways, depending on the size and type of energy source available and the water quality product requirements. There are economic and technical advantages of hybrid as compared to single process technology.

These include the utilisation of a common seawater intake, optimised feedwater temperature for the RO plant, taking cooling water from MSF or MED plant, blending of product waters, common water treatments and various other optimisations that can be made through common process requirements. Some of the different hybrid desalination systems are reviewed in (Awerbuch, 1997).

Some of the reactor concepts that are under consideration for desalination applications are shown in Table 14.5 and discussed below.