The monolithic raft footings used in building nuclear power plants largely use large bored piles up to 1.50 m or so in diameter, with piles approx. 4-6 m apart. One particular feature arises if bituminous sealing is used. In that case, the slab on ground is often divided into two separate slabs: the lower pile head slab and the upper building seal slab, between which the sealant is then applied.

Bored piles are often used in foundation work in existing structures, as they are largely vibration-free.

When performing foundation structures in existing buildings, work often has to be done close to, or even over, safety-related underfloor structures. There is usually not much space for large drilled piles in such areas, so that micro-piles or continuous flight auger piles are used, which can be used virtually vibration-free even in the vicinity of safety-related pipes.

Particularly in the vicinity of safety-related pipes in the subsoil, particular attention must be paid to the horizontal loads transmitted by wind, earthquake etc. to avoid putting any additional stresses on the pipes. This can be avoided, for example, by ‘insulating’ piles in the vicinity of pipes or using angled piles if there is enough space.

5.2.3 Strength parameters

In principle, the strength parameters for concrete, reinforced and pre-stressed concrete, including limits of ultimate limit strains to be observed, must be taken as per DIN 1045­1 [54]. Further details specific to nuclear power plants can be found in DIN 25449 [15].

To determine the design values, the characteristic strength parameters must be divided by the partial safety factor gM in each case. Concrete compression strength fc must be divided by gc and the strength of concrete steel and pre-stressing steel (concrete steel: yield stress fyk and tensile strength ftk, pre-stressing steel: yield point fp0.1k and tensile strength fpk) must be divided by gS, using the partial safety factors as shown in Table 6.2 for the proofs in ULS depending on the category of requirements concerned.

For concrete compression strength, the influence of long-term effects and the influences on design-relevant concrete characteristics, such as the effects of load duration, curing and loading speed, must be taken into account. In certain justified cases, variations from the design-relevant characteristics of concrete as a construction material from the character­istics on which DIN 1045-1 is based may be used (deviations from design values). This applies in particular to the concrete getting stronger as it cures in long-standing reinforced concrete structures and the increase in strength of concrete stressed in multiple axes or high expansion rates and the influence on the ultimate limit strains of concrete.

To provide permanent protection, structural waterproofing not only must it be designed professionally and quality assured, but it must also be built accordingly. When building the structure, the external conditions involved while making the structural waterproof play a governing role.

The issues of importance in connection with building the structure, that make a major contribution to quality assurance (as below), are:

— particular measures to be taken if constructing in bad weather

— temporary protection measures:

— protecting the structural waterproof on horizontal and gently sloping surfaces

— protecting waterproof connections

— protecting against thermal effects

— protecting the structural waterproof on wall surfaces when placing reinforcement

— protecting against penetrating groundwater, accumulating and surface water during construction

— protecting against the penetration of harmful substances.

Most elements on Earth are stable, and the structure of their atomic nuclei is constant. A few of them decompose radioactively, however: that is to say, their atomic nuclei turn into those of other elements by emitting radiation or particles.

In a nuclear reactor, or a reactor at a nuclear power plant, nuclear fission is induced deliberately and the resulting radioactive decay used. Atomic nuclei are split by bombarding them with neutrons.

Design and Construction of Nuclear Power Plants. First Edition.

Rudiger Meiswinkel, Julian Meyer, Jurgen Schnell.

© 2013 Ernst & Sohn GmbH & Co. KG. Published 2013 by Ernst & Sohn GmbH & Co. KG.

Plutonium 239,

Other transuramc e ements

Fig. 2.2 The nuclear fission process

The process of nuclear fission is shown in Figure 2.2. In the reactor, uranium U-235 nuclei are bombarded with neutrons, causing them to fission and emit radiation, known as ‘nuclear radiation’ (cf. Section 2.3). The products of decay are usually two fission products, such as krypton or barium, and two or three neutrons. The neutrons that are emitted can in turn split other atomic nuclei, setting off a chain reaction in which energy is released.

The fission products that arise when atomic nuclei split are unstable: they give off radioactive radiation, turning into stable end products, releasing more energy in the process. This post-decay heat keeps on being generated even after a nuclear reactor has been shut down, and requires special post-cooling systems (Figure 2.3).

A constant steady chain reaction needs a certain minimum mass of fissionable material, also known as the ‘critical mass’. Critical mass exists if the number of secondary fissions (second generation neutrons) is equal to the number of primary fissions (first generation neutrons).

Uranium U-235 is the only element occurring in nature that can maintain fission via a chain reaction. U-235 accounts for just 0.72% of the total mass of uranium occurring

naturally, so it does not provide the critical mass required: this has to be increased, i. e. the uranium has to be enriched. This can be done using diffusion, gas centrifuges or separation nozzles.

The critical mass of U-235 required is less if the neutrons that are released when its nuclei split can be slowed down to lower, thermal speeds (moderated). This can be done using what is known as a moderator. Apart from carbon in graphite form and heavy water (deuterium oxide, or D2O), this is best done using light water, or H2O. The water molecules slow the neutrons down very effectively, thus maintaining the chain reaction; and the water absorbs the energy from nuclear fission, which heats it up considerably, making it ideal for generating electricity. When using H2O as moderator, the natural uranium has to be enriched to around 3.5% U-235.

The large number embedded parts involved (a nuclear power plant block may have more than 100,000 anchor plates) calls for a particular feature of planning, such as recognising collisions in good time and avoiding them, and particular preparations on site to ensure that they can be finished on time in parallel with the formwork and reinforcement work. As well as the anchor plates just mentioned, fitted components include such items as pipes, foundation frames and the frames for Omega water stops, known as Omega frames. These are attached to the formwork or to special support structures, and this must be carried out in such a way as to maintain the tight tolerances in terms of precision location once concreting is complete.

With the OL3 construction project, the anchor plates used to fix components later on are made largely of ferritic steel anchored by headed studs (Figure 4.23).

These anchor plates arrive on site coated with rust-protection base coat, and are painted in the finishing phase. The plates are painted once again once the load-bearing structure is in place.

Pipe lead-throughs of ferritic or austenitic steel are installed in the first — or second-cast concrete. Fitting them at the second-cast concrete stage means an extra work process before handing over to the mechanical trades, which could delay the latter starting; but the installation quality is generally higher in terms of precision.

7.1.2 History

The use of headed studs in fastening systems dates back to the early 1970s. The first headed studs used in building nuclear power plants were PECO concrete anchors. In 1971, Peco Bolzenschweifitechnik GmbH was acquired by Nelson StudWelding Co. and had its name changed to Nelson Bolzenschweiss-Technik GmbH. After a long time as part of the TRW-group, Nelson Bolzenschweiss-Technik moved to the Fabri-Steel Group in Michigan, USA, in 2000, and has belonged to Doncasters Group Ltd, UK, since 2009. However, the company label, Nelson, remains unchanged to date.

The headed studs found in older nuclear power plants under the names ‘Peco’, ‘Nelson’ and ‘TRW-Nelson’ are all the same product. Following extensive tests and expert opinions by Professors Roik and Bode, the DIBt granted the first general approval for headed studs (anchoring steel plates using welded-on Nelson headed studs, approval no. Z-21.5-82) in 1983.

Not long after that, the DIBt issued further headed stud licences for the Koster & Co. of Ennepetal (Z-21.5-280) and Riss AG of Dallikon, Switzerland (Z-21.5-296).

Headed stud anchors were still designed based on permissible tension and lateral loads at the time. Existing edge and plane factors were allowed for via reduction factors (kappa method). When the licence was amended in 1995, an extended calculation method (CC method) was introduced and the semi-probabilistic safety concept was used instead.

The DIBt issued the first European Technical Approval for fastenings in November 2003. The two licence notices ETA-03/0039 (KOCO headed studs) and ETA-03/0041 (Nelson headed studs) were extended in 2008, and are the current state-of-the-art rules.

The particular safety requirements involved that mean using headed studs in nuclear power plants must be approved by the licensing authorities in each case.

The measures required for ageing management purposes depend on the maintenance strategy implemented (cf. Figure 9.2). As well as the maintenance work itself, these include the inspections with regular tests and special tests, repairs and strengthening in the event of structural changes.

The basic procedure for conducting ageing management to KTA 1403 is shown in Figure 9.3 (PDCA cycle). Ageing management is knowledge-based, incorporating all information on structural systems and being constantly updated.

Inspections and structural examinations are used to establish what condition each structure is in, and to detect faults and changes. They are conducted at regular intervals in accordance with specified structural lists and the results recorded in status reports, so that trends can be followed in status reports and any faults observed can be assessed and repairs made.

Structural examination methods are laid down in instructions. Each examination is normally followed by visual checks, simple measurements or as part of geodetic survey programmes, using preset criteria to specify more intensive examination methods.

In terms of reporting, KTA 1403 requires specific system basic reports, annual status reports and building status reports for structural systems (every 10 years). The basic report describes the process of ageing management, including organisation, and includes ageing relevant findings. Status reports contain details of ageing relevant activities and measures, findings and results during the reporting period. Building status reports are designed to show that safety-related structures have been assessed to see how they have aged.

Basic and status reports, which both cover mechanical engineering, electrical engineer­ing and structural engineering, may also include individual specialist reports. For structural systems, for example, a ‘Basic report — structural engineering’ and ‘Status report — structural engineering’ (including building status report issues), can be prepared.

3.2 General notes

Nuclear plants are divided into generator reactors, research and training reactors and nuclear fuel supply and disposal systems. This includes, in particular, nuclear power plants for generating electricity, fuel element production installations, uranium enrich­ment plants, protective structures such as ponds and storage facilities for radioactive waste, which in turn are divided into interim and final storage facilities.

Building structures required for nuclear plants whose protective function means that they are classified as safety-related (cf. Section 2.5) have to meet particular construc­tion requirements. These requirements, which are more stringent than those involved in conventional construction, must be observed not just when designing and constructing buildings but also when operating and dismantling nuclear plants.

For earthquake design purposes, KTA 2201.1 [37] divides components and building structures into three classes, as follows:

— Class I

Components and building structures that are required to fulfil the protective goals (control radioactivity, cool fuel elements and contain radioactive substances) and limiting radiation exposure (safety-related system components and building structures)

— Class Ila

Components and building structures that do not belong to Class I, but which, due to their own damage and the sequential effects, possibly caused by an earthquake, could detrimentally affect the safety-related functions of Class I components and building structures

— Class Ilb

All other components and building structures

The only components and building structures for which seismic safety is required are those in Classes I and IIa. Components and building structures of Class I must be verified in terms of load-carrying capacity, integrity and functional capability, i. e. deformation or crack widths in reinforced concrete must be limited in some cases. For components and building structures of Class IIa, generally verification of load-carrying capacity will be sufficient.

To verify earthquake safety, structural analyses are required reflecting the design basis earthquake and its possible consequences. Possible consequences could include the failure of high-energy containers, not designed to withstand earthquakes, such as feed water tanks in the turbine building of a PWR plant. Combined effects of earthquakes and other extraordinary actions are not generally taken into account as they are extremely rare.

For structural analysis purposes, earthquake effects are to be set as the ground response spectra for the reference earthquake or compatible recorded acceleration over time curves in each case, recording the simultaneous excitation in both horizontal and the vertical direction. The subsequent superposition of parallel stress variables can be taken either as the root of the sum of the quadratics or the superposition rules as in DIN 4149 [39] or DIN EN 1998 [40].

Structural modelling is subject to particular requirements, due to the dynamic effects and to the influence of the subsoil at the site in particular. Precise details of structural modelling, including details of structural damping and subsoil modelling can be found in KTA 2201.1 [37], KTA 2201.2 [41], KTA 2201.3 [42] and KTA 2201.4 [43].

In principle, the structural models to be used for the building structure, including the subsoil for the plant components with their support structures are those which record how the structures behave in the governing frequency range of an earthquake. Depending on the purpose of verification involved, it must be decided whether structural modelling requires a level beam model or a spatial beam model or even a spatial surface structure model, allowing for possible decouplings between the building structure as a whole and part structures or decoupling criteria between the building structure as a whole and components.

As far as the dynamic behaviour of the structure is concerned, the influence of the interaction between structure and subsoil (subsoil-structure interaction) must be taken
into account, varying the soil characteristics to give a lower, medium and upper subsoil strength. The results of the calculations at different subsoil strengths must then be included.

The structural analyses can be carried out using the usual dynamic calculation methods, including in particular the response spectrum method, frequency range method, time history method and the quasi-static method as a simplified method. These are generally used as linear methods. Non-linear methods such as non-linear time history methods are also used in exceptional cases.

The result of the dynamic structural analyses, as well as eigenfrequencies, is to give the internal forces and deformation variables required to assess the strength and deformation behaviour of the structure studied. Response spectra can also be calculated at the intersections with other building structures or components to use these to analyse the building structures or components meeting at these nodes. The resulting method to be used in conducting structural analyses of building structures and components with a view to using response spectra is therefore as follows (cf. Figure 5.5):

— Specify the site excitation as ground response spectra or time history (primary response/primary spectra)

— Calculate the response over time or response spectra of the structure (secondary response/secondary spectra)

— Calculate the response over time or response spectra for system components (tertiary response/spectra)

Ground response spectrum

Frequency

Tertiary spectrum

Fig. 5.5 Response spectrum method (building structure/components)

5.3.2 Floods

Issue 9/98 of the Deutsches Institut fur Bautechnik’s Guideline on ‘Using anchors in nuclear power plants and nuclear installations’ [63] constitutes for the first time what tests are required over and above the General Approval in order to facilitate use of anchors for safety-related attachments in nuclear power plants.

This Guideline contains details of the tests needed to simulate the extraordinary stress situations involved in an earthquake. To prove their suitability, anchors must withstand a monotonic tension load, alternating loads at constant crack widths and varying crack widths at constant loads at a crack width of 1.5 mm. These test loading conditions simulate comprehensively the accidental stresses due to earthquake actions. To obtain the characteristic tension loads, monotonic tensile tests are conducted at an open crack width of 1.0 mm. The characteristic shear strength is determined by alternating shear load tests, in which anchors are exposed to 15 times the alternating shear loads in the direction of the crack. The residual load-bearing strength is then determined by a monotonic shear tension test.

The new edition of the Guideline of June 2010 [67] revised the crack widths. This also allows for the specific crack widths that are expected at the place of use to be used in tests. However, this requires a detailed verification of the characteristic crack widths under accidental actions. An additional section has also been added on testing to determine realistic anchor shifts. Under the DIBt Guideline, anchor shifts are determined by tests in opening and closing cracks at a constant tensile load acting on the anchor and tests at alternating loads on the anchor with the crack opened, varying the crack widths as well as variable alternations of crack widths and loads. To pass these tests, anchors must satisfy a numbers of requirement defined in the DIBt Guideline.