When establishing the basic design concept for a white tank, there are a number of local conditions to be considered:

— height of the reference groundwater level (and/or height of accumulating seepage water)

— height of the minimum groundwater level before starting use

— chemical composition of the groundwater.

If there are any liquid factors to be dealt with from inside, the following must also be established:

— maximum possible height of the liquid level surface

— chemical composition of the liquid

— maximum possible temperature of the liquid

— duration of the liquid load.

Particular attention must be paid to limiting any water or other liquids escaping in the event of an incident.

Establishing categories of use under the WU guidelines cannot be applied directly to nuclear power plants, particularly because of what is required in the event of an incident. Instead, unique findings must be made for each individual structure, broken down by structural components, such as floor slab, walls and so on, if possible.

In terms of groundwater loads from outside, we need to define to what extent limited local access of groundwater can be accepted:

— in the phase of structural works, shell and core

— before commissioning

— in use

— during and after incidents.

Requirements for liquid stresses from inside must be defined analogously.

Except in the event of an incident, external structures should be designed and dimensioned in accordance with the DAfStb guidelines on concrete structures when dealing with water pollutants [102].

Design principles

If separating cracks arise in WU structures, these can admit large quantities of water [103]. Even fine cracks usually admit more water than can be displaced by air from the inside. How much water gets in depends mainly on how wide a crack is, how thick a structure is and how high the water level is.

When dealing with separating cracks, the design principles which can be used for nuclear power plant structures exposed to standing groundwater are:

— avoiding separating cracks

— designing to waterproof cracks on schedule prior to commissioning

— cracks waterproofing through self-healing

— designing to run off penetrating water.

Whether individual design principles can be implemented must be considered in the light of the construction timetable.

The design principles selected must be justified and recorded, and the findings made must be included in contracts between the parties involved.

The water which is typically used as coolant in light water reactors can be used both in a single-circuit system or — to prevent contamination — in a multiple-circuit system via heat exchangers. Light water reactors are known as pressurised water reactors (PWRs) or boiling water reactors (BWRs), depending on whether the water in them is pressurised or boiling.

In a pressurised water reactor, the water in the reactor pressure vessels is at extremely high pressure, around 150 bar, so the water does not boil, even at the design operating

6.

Generator 11. Cooling water treatment

7. Transformer plant

8. Condenser 12. Cooling water pump

9. Heat exchanger 13. Power station basin

10. Feed-water pump 14. Cooling tower

Fig. 2.7 Nuclear power plant with pressurised water reactor (PWR) temperature of 300 °C. This prevents steam bubbles forming, which would complicate the heat transfer process.

As Figure 2.7 shows, a PWR has two coolant circuits: the primary and secondary circuits (water-steam circuit). In the primary circuit, the coolant water flows round the fuel elements directly: the water which is heated in the reactor core of the reactor pressure vessel is then fed to the boiler and back to the reactor core via circulating pumps. The steam generator then transmits the heat to the secondary circuit, producing steam which drives the turbine and consequently the generator, so the steam passing to the turbine is not radioactive. At the end of the secondary circuit, the steam which was depressurised and condensed in the condenser (coolant water circuit) is pumped back to the steam generator via heat exchangers (preheater unit).

Unlike a pressurised water reactor, in a boiling water reactor (Figure 2.8), the water in the reactor core of the reactor pressure vessel is heated to boiling point: so compara­tively little pressure is required at the proposed operating temperature of 300 °C. A pressure of 70 bar is sufficient. Nor is a boiler required, so only one coolant circuit (direct circuit) is necessary. The live steam is fed directly from the reactor pressure vessel to the turbine, which means that the turbine becomes radioactively contaminated to a limited extent. Unlike with the PWR, in which the reactor is controlled and can be crash shut down by control rods from above, with the BWR, control rods are inserted into the reactor core from below. (Please note: control rods are used to control and shut down nuclear reactors.)

If a loss of coolant accident (LOCA) occurs (Section 2.5) in a BWR the pressure is reduced by condensing the steam released in a condensation chamber, so the safety

1.

Reactor pressure vessel

4. Turbine set

11. Cooling water pump

12. Power station basin

13. Cooling tower

Fig. 2.8 Nuclear power plant with boiling water reactor (BWR) vessel containing the reactor pressure vessel in the reactor building is much smaller than for a PWR of comparable output.

Of the eleven PWRs and six BWRs operating in Germany, the three PWRs of the Convoy model (Siemens KWU) with an output of approx. 1400 MW are the most advanced. One of these Convoy plants, which may be classified as Generation III, operates at the Isar site (near Landshut) together with a BWR unit (Figure 2.9).

Fig. 2.10 Overall view European pressurised water reactor EPR (AREVA, 3D visualisation)

In the course of the further development of the Convoy nuclear power plant design, German nuclear power plant operating companies decided to join forces with the French state company EDF to develop the EPR (European Pressurised Water Reactor) in 1991. This EPR, a Generation III+ model generating 1600 MW, is currently being built in Finland and France (Figure 2.10). It is being supplied by French plant supplier AREVA, which acquired the former Siemens KWU some years ago.

As well as EPR, AREVA with German involvement (E. ON Kernkraft) is also developing the boiling water reactor KERENA (formerly designated SWR 1000) with an output of 1250 MW (Figure 2.11).

Further new developments in Generation III+, which are now being offered and preferred as large-scale power plants with outputs of well over 1000 MW each, are the boiling and pressurised water reactors as listed below (Figures 2.12 and 2.13):

— ABWR: BWR — 1350 MW;

Supplied by: Westinghouse (USA)/Toshiba (Japan),

— AP1000: PWR — 1000 MW;

Supplied by: Westinghouse (USA),

— AES 92: PWR — 1000 MW;

Supplied by: ASE (Russia),

— APR1400: PWR — 1400 MW;

— Supplied by: KOPEC (South Korea).

There are basically two decommissioning options to choose from when dismantling nuclear power plants:

— Dismantle immediately, as soon as the rundown phase has been completed

— Safe confinement: after the rundown phase, put the nuclear power plant into ‘safe confinement’ for around 30 years before starting to demolish it

Immediate dismantling

— Demolish all contaminated and active building sections, systems and components immediately

— Prepare and pulverise all radioactive waste for interim or final storage

— Decontaminate and release other remains

— Decontaminate and release building, demolish conventionally

Safe confinement

— Demolish all contaminated structures, systems and components outside the con­tainment area immediately

— Reduce control area and prepare and pack radioactive waste involved for interim or final storage.

— Decontaminate and clear other residuals involved

— Clear media (press, TV etc.) if possible

— Leave active structural components (nuclear installations, pressure vessel, bioshield) as installed, seal system interfaces appropriately

— Continue to operate essential systems during safe confinement (ventilation, pres — surisation, monitoring systems)

— Confine safely for 25-30 years

— Apply for dismantling permit during safe confinement phase (around five years before safe confinement ends)

— Create new infrastructure facilities

— Demolish and clear plant as with immediate demolition

— Timescale:

Establish safe confinement: 5-8 years Operate in safe confinement mode: 25-30 years Demolish completely: 8-10 years

In Germany, the only nuclear power plants that have been put into safe confinement are Lingen (KWL) and Hamm-Uentrop (THTR). In the light of experience gathered with dismantling projects to date, the prevailing view today is that starting dismantling as soon as the rundown phase is complete is preferable. The advantages include: the nominal costs of direct dismantling are less than those of safe confinement, plant personnel are still on hand, personnel can continue to be employed and the site can be available to be reused sooner if required. The considerations in favour of safe confinement include reducing potential activity in the plant from radioactive decay, possibly using technical innovations and developments and reducing immediate costs.

7.1.3.1 Basics

During the construction period of the first nuclear power plants no general authoritative approvals for headed studs had existed. Anchorings were designed and built based on consent orders from the building authorities concerned. These orders and their associated expert opinions governed the permissible combinations of stresses for tension and shear loads for headed studs and covered any boundary or group influences individually.

When designing anchor plates, a distinction is made between safety-related and non­safety-related units.

The design and construction of fastenings for non-safety-related components is based today on the general technical approval (ETA) for the significant combination of characteristic loads concerned.

Beyond the provisions of the General Technical Approvals, safety-related components are subject to additional requirements in terms of the structural load-bearing and deformation behaviour of the anchorings.

As well as for the effects of actions from characteristic loads, safety-related anchorings with headed studs must also be designed to withstand the effects of actions from accidental external events such as earthquakes or for internal anomalies. Designing anchorings for special load cases also includes the effects of impulsive actions and the occurrence of cracking in reinforced concrete structure with wide cracks.

The design of anchor plates within headed studs for attaching safety-related compo­nents is not covered by General Technical Approvals; nor has any such General Technical Approval for accidental actions (K-approval) yet been applied for at the Deutsches Institut fuur Bautechnik. That means using anchor plates for safety-related components is still subject to approval by the supreme building authority in each case. This involves verifying structural strength and serviceability of anchor plates in accordance with DIBt Guidelines [67], the successor to [63]. Finally, the suitability of the anchoring for the specific intended purpose in nuclear power plants must then be assessed by an expert opinion.

When assessing shielding levels, a distinction must usually be made between gamma radiation and neutron radiation. DIN 25413 [18] classifies shielding concretes by the proportion of elements they contain. How much shielding concrete provides against gamma rays depends directly on the bulk density of the concrete and the proportions of elements it contains by weight. In other words, the higher the weight of concrete, the more shielding it provides. With neutron radiation, how much shielding concrete provides depends on what chemical elements it contains. As well as using additives with crystal water content, the proportion of light elements, such as hydrogen, is particularly important, (or boron compounds are also used, such as boron carbide, borax frit, colemanite, boron calcite) which are particularly good at ‘trapping’ fast neutrons.

Raw density specifications are generally based on dry weight or dry raw density. DIN 25413 defines different compositions of concrete mixes and the main element proportions involved, such as O, C, Si, Ca and Al or it recommends a so-called average composition. For heavyweight concrete, this standard also specifies different kinds of concrete and characteristic proportions of elements, depending on the heavy aggregates used (haematite, magnetite, ilmenite, barytes, limonite and serpentine). Under DIN EN 206-1 [19] and/or DIN 1045-2 [20], heavyweight concrete has a dry specific gravity in excess of 2.6 t/m3. However, DIN 25413 refers to an older definition of heavyweight concrete. Radiation-proof concrete made with heavyweight aggregates therefore generally has a raw density over 2.8 t/m3.

Heavyweight concrete is considerably more expensive to bring in than standard concrete. Because of the largely angular aggregates and higher density involved, it does not pour nearly as well as standard concrete, and it requires more careful mixing to ensure that components of different density do not separate. For notes on this, and an overview of heavy aggregates, see the DBV code of practice for radiation protection concretes [21].

Heavyweight concrete as radiation protection concrete was already being used in the first nuclear facilities in Germany in the 1960s, at the Jiilich research centre (research reactors DIDO and MERLIN). So-called ball scrap concrete (with cast-iron granulate), for example, with a specific gravity of 5.6 t/m3, had been used.

In more recent plants, however, additives such as magnetite, serpentine, haematite or barytes or in some cases granulated iron additives had been used, as they are easier to work.

A cement content of 340kg/m3 (CEM III/B 32.5), a water content of 170kg/m3, with 1410 kg/m3 of haematite 0/6, with 1680 kg/m3 haematite 6/25 and 150kg/m3 sand 0/8 has been used to give a specific gravity of 3.6 t/m3, for example. Using 890 kg/m3 haematite 0/8 instead of sand 0/8 and 1920 kg/m3 haematite and an extra 1350 kg/m3 iron granulate can give a bulk density of approx. 4.5 t/m3.

Heavyweight concrete as radiation-proof concrete is mostly required in the immediate environment of the reactor pressure vessel as part of the bioshield. In the support area of the reactor pressure vessel (known as the skirt area) of the bioshield in the containment at Gundremmingen a heavyweight concrete with specific bulk densities of 2.7-4.2 t/m3 was used, for example.

What is particularly important in construction terms is the dry specific bulk density of a normal concrete to be designed for radiation protection purposes. Using normal quartz gravel as aggregate can only reliably give a dry specific gravity of 2.2 t/m3. If shielding requires a higher specific bulk density, it must be borne in mind that special aggregates will be required. These may have to be brought by considerable distances, which could increase costs.

KTA 2207 [23] defines the reference water level for coastal sites and sites on tidal flows as a storm tide water level with an exceedance probability of 10~4/a. This storm tide water level SFWH(10-4) can be obtained using suitable but highly laborious probabi­listic methods, which can also be used to determine flood runoffs (cf. [45]). Alterna­tively, according to the annexe to KTA 2207, a probabilistic based extrapolation method can be used, taking the storm tide water level SFWH(10-4) as the total of a base value BHW(10-2j and an extrapolation difference ED as follows:

SFWH(10-4} = BHW(10-2) + ED.

The design basis water level BHW(10 2) with an exceedance probability of 10~2/a is calculated here based on a quantitative statistical extreme value analysis. The spread of the results with the usually long, good-quality time series of water levels on the coasts and in tidal flows is relatively low.

Determining the extrapolation difference ED calls for detailed studies of the coastal or estuary levels of the tide flows concerned. At the water gauge sites of Cuxhaven and

Brokdorf on the river Elbe, for example, this gives an extrapolation difference ED of the order of 100-150 cm.

With dykes, as well as the storm tide water level SFWH(10-4) the wave run-up must also be taken into account (Figure 5.6) and, having superposed these two variables, the dykes must be designed without waves breaching them or a possible breaching wave putting the stability of the dyke at risk. The wave run-up height at the dyke depends not only on the wave height and wave period, but also on the characteristics of the dyke itself, such as its slope or surface area. When calculating wave heights, it should be borne in mind that these are particularly subject to local wind speed and direction and to the topography of the foreshore.

Anchor plates with headed studs or metallic anchors consist of non-flammable materials and can therefore be assigned to fire protection class A1 under DIN EN 13501 [91]. Should anchor plates be subject to specific requirements in terms of fire resistance time, structures must be tested in accordance with the test procedure prescribed for their class and their fire resistance class is to be specified according to DIN EN 13501.

Safety requirements for structural systems can be deduced overall from the statutory requirement to prevent damage and from the safety goals to be complied with.

Specific requirements here are laid down in the nuclear rules, accident rules and KTA safety standards.

The storage building is required mainly to:

— provide shielding

— remove heat

— be designed for operating and exceptional loads

— provide protection against fire and lightning strike

— protect against the weather

— protect against third parties (sabotage).

4.3.2.1 Design criteria

Design criteria are governed by:

— Shielding

Most of the ionising radiation that fuel elements emit is shielded by their containers. The reinforced concrete building structure provides further shielding, keeping radiation levels within the limits laid down by the radiation protection regulations and protecting staff and the environment.

— Heat removal

The interim storage facility design is designed to remove the heat that the fuel elements give off as they decay, by way of natural convection. The air inlets and outlets required must be arranged and dimensioned to remove heat reliably.

— Building settlement

Building settlement due to the container loads involved must not compromise the structure or the operation of the cranes etc. Settlement is estimated technically at the planning phase, allowing for subsequent partial occupation levels, and is monitored in operation via recurrent settlement testing.

— Structural integrity

As with conventional structures, this requirement can be met via the rules of building design on the design of the roof and sealing the building externally, if groundwater conditions allow.

— Floor structure and decontaminatable coatings

The slab and ground in the storage area must have sufficient compression strength and wear resistance to take the containers put into storage. This is achieved by using a mechanically smoothed concrete surface with hardening agents mixed in. In the reception and maintenance area, the floor is given a decontaminatable coating as a precaution. In the loading and unloading zone in the reception area a shock-absorbent layer of so-called damper concrete can be included in the floor slab to protect containers and floor slab if a container is dropped from a height of 3 m, which cannot be ruled out.

— Durability

Interim storage facilities are designed to be permanent in accordance with conventional standards. If they are built properly of tried and tested reinforced concrete designs, they should last for their full working lives.

4.3.2.2 Building design

As we saw in Section 4.3.2, building structures in Germany fall into one of two different designs: WTI and STEAG. These designs differ from one another in particular in terms of their structural design.

WTI Design

The building is designed to withstand exceptional effects from outside, such as earthquakes and blast waves from explosions. They do not need to be designed to absorb aircraft impact, as the containers themselves are designed for this external event.

Exceptional events from inside are containers falling from a height of 0.25 m in the hall area and 3.00 m in the loading area. In the trans-shipment hall, so-called damper concrete is used in areas in which containers could fall, to absorb the energy released, enabling the loads involved to be transmitted without additional strengthening the floor slab at this point.

When floor slabs are occupied by CASTOR containers in blocks of eight, this gives a floor slab loading of 200kN/m2.

What is not typical, compared with similar lightweight hall constructions, however, is the roof construction; this has to be 55 cm of normal concrete to be radiation-proof. This high permanent load component means that this design has to have relatively high roof girder constructions.

STEAG Design

The greater roof slab and wall thicknesses of the solid STEAG design will at least protect against penetration from aircraft impact. Unlike the WTI design, such halls can also hold containers designed for a debris load of 21 at least, should roof sections fall in.

Temperature effects

The relatively high room temperatures of approx. 80 °C mean that the outer walls and roofs must be reinforced accordingly, to guard against a correspondingly high crack width, which must be demonstrated in many areas for centric forces as finally built (with the concrete at its full tensile strength).

The floor slabs are designed not merely for a high load per surface area of up to 200kN/m2, but also for hot spot temperature effects of approx. 120 °C immediately below the containers. This makes an additional consideration of the upper reinforcement of the floor slabs necessary. The hot areas cause concentric inherent stresses leading to cracking.

However, non-linear studies of floor slabs made at interim storage facilities have shown that no additional reinforcement is required because of the hot spot effect.

5.2.4 Requirements of containments

The containment (safety container or enclosure) in the reactor building of a nuclear power plant is the essential structural barrier involved in containing radioactive

substances safely (cf. Sections 2.5 and 4.2). The verifications required for this barrier are:

— bearing capacity

— serviceability in the sense of functional ability

— integrity (gas-tightness).

Bearing capacity and serviceability for use can be combined as a single overall concept, structural integrity. The structural integrity of a containment is tested once it has been made, while gas-tightness is tested regularly every three to five years.

The verifications must take account of the actions when operated as intended (normal and abnormal operation) and those of incidents (cf. Section 2.5). Containment design is governed in particular by the possibility of a loss of coolant accident, with its high pressure of the order of 0.5 MPa accompanied by temperatures of approx. 150 °C.

The inside surfaces of white tanks must be kept permanently free for inspection purposes and to carry out any waterproofing which becomes necessary. Should they only be accessible by dismantling replacement parts, equipment and so on, this must be capable of being done with operations on the run. If access cannot be assured in the area of floor slabs under machinery and so on, structural design measures must be taken to ensure that, if any water does penetrate, it cannot cause damage and can be run off as designed. Wall claddings on the inside of WU structures (liners and tiles) are not allowed.