As with the verifications for shear resistance, punching shear verifications for reinforced and pre-stressed concrete structures in nuclear installations must be conducted to DIN 25449 [15]. Under these verifications, which are also based on DIN 1045-1 [54], punching shear reinforcement is required if, along the critical circular section to DIN 1045-1, the shear force vEd to be absorbed per unit of length is greater than the shear resistance vRd, i. e. if:

vEd > vRd;ct

Shear resistance vRdct is obtained from DIN 1045-1. Like VRdct, vRdct reflects the different requirement categories A1, A2 and A3 by a factor (cf. Section 6.3.2: 1.0 for A1, 1.15 for A2 and 1.50 for A3).

To find the punching shear reinforcement required, we distinguish between

— structural members subject to indirect effects of action, as covered in DIN1045-1 (e. g. supports in slabs or foundations), and

— structural members subject to direct effects of actions, as they occur in nuclear engineering construction as structural members subject to extraordinary actions in requirement categories A2 or A3, such as airplane crash or jet forces.

The reinforcement of structural members subject to indirect effects of actions must be obtained to DIN 1045-1. Should no more precise calculation and design procedure be used, the reinforcement required for structural members subject to direct effects of actions may be calculated in accordance with DIN 25449, which is based on DIN 1045­1 and additional experimental studies. The upper bound vRd, max to DIN 1045-1, which is intended to prevent the concrete cover palling at column faces, is irrelevant here, although the concrete strut resistance must be verified to ensure that the stirrup reinforcement is activated, i. e.

VEd ^ VRd, max = °-25 ‘ fcd ‘ uload ‘ d (6.11)

Where

fcd is the design value of concrete compression strength [N/mm2]; fcd = fck/gc (gc in

Table 6.2)

uioad is the circumference of the load area Aload (equivalent circle with radius Rioad; see Figure 6.1)

d is the static effective depth of the flexural reinforcement of the side facing away from the load in the cross-section considered

In calculating the punching shear reinforcement, the decisive shear force VEd is taken as the maximum load resulting on the load area Aload. The verification may assume as the failure figure a punching shear cone with an effective surface area Acw = p x (rL — rn) and angle of inclination of the punching shear cone fir (generally cot fir = 1.25) (Figure 6.1).

For the reinforcement, it must be shown that the relationship

VEd < VRd, sy (6.12)

is satisfied.

The design value for the required shear force reinforcement VRd, sy consists of a contribution of concrete load-bearing VRdc (with the contribution of longitudinal reinforcement) and a contribution of punching shear reinforcement, i. e.

VRd, sy VRd, c T ks ‘ Asw ‘ ftd (6 13)

= [cd • 0.17 • k • (100 • P!)1/4 • f^k^3] • d • uex + asw • ks • Acw • ftd

Where

(average useful height in mm) cd Prefactor reflecting requirement category

Pi Degree of longitudinal reinforcement in tension area

7.3.1.1 General requirements

Reinforced concrete structures with high resistance to water penetrating prevent water penetrating permanently in liquid form. As well as bearing loads, they also act as structural waterproofing.

Designing and constructing white tanks in Germany is governed by the DAfStb’s guidelines on ‘Water-impermeable concrete structures’, (WU guidelines) [98], as the generally accepted rules of the art. The WU guidelines provide instructions on requirements for fitness for use of water-impermeable reinforced concrete structures. DAfStb vol. 555 ‘Explanatory notes to the DAfStb guidelines’ [99] contains notes to the WU guidelines. Instructions are contained in DBV bulletin on ‘High-grade use of basement floors — building physics and indoor climate’ [100].

There are a number of constraints to be considered when building a white tank:

— nature of the moisture penetration

— design water levels

— type, characteristics and permeability of the subsoil

— chemical characteristics of the water

— establishing the type of use involved, particularly in the light of the stresses and special loads from external and internal events.

Supplementary drainage measures may be taken to protect the structure, particularly against non-pressured water or temporarily accumulating water. For drainage design, dimensioning and execution rules, see DIN 4095 [101].

1.1.1 Overview

Many kinds of nuclear reactors have been developed since the discovery of uranium’s nuclear decay in 1938. These can be divided into generations, in the order in which they were developed, as follows:

— Generation I

The initial prototypes built between 1957 and 1963.

— Generation II

Commercially viable reactors built from the mid 1960s onwards.

— Generation III

Advanced reactors, generating much more power and with much more concern about safety, built since the early 1980s.

— Generation III+

The next generation of reactors, with structural safeguards against meltdown and/or passive safety features.

— Generation IV

The reactors of the future, highly efficient, with advanced safety features and producing little spent nuclear fuel, but not expected to come on stream until 2030 at the earliest.

(Remark: At the international level, Generation III is often classed as part of Generation II, so Generation III+ is referred to as Generation III.)

Of the types of nuclear reactor that have been developed, there are only a few that can be used in commercial operation. The different types can be broken down by the following aspects:

— Fuel

e. g. natural uranium, enriched uranium, plutonium, thorium; whether they use clad or unclad solid fuels (cladding materials are zirconium, aluminium, magnesium or magnesium oxide — Magnox); fuel elements may be rods, plates, tubes or pellets

— Neutron energy

thermal reactors (moderated neutrons, using moderators such as graphite, light water H2O or heavy water D2O) and fast reactors (without moderating the neutrons)

— Coolant

light water H2O, heavy water D2O, gas (air, but mainly carbon dioxide and helium).

The first basic distinction here is between thermal and fast reactors. Fast reactors are better known as fast breeders, because when they are operating they ‘breed’ more fissionable plutonium from the uranium than they use, which means that they can get around twice as much energy out of the uranium. Fast breeders have failed to establish themselves, however, for a number of reasons (political reasons in Germany).

Amongst the thermal reactors, there are a number of combinations of moderators and coolants which have been developed successfully for commercial use (Table 2.1). The two main families involved here are gas-cooled reactors (Magnox reactors), advanced gas-cooled reactors and high-temperature reactors and water reactors (light and heavy water reactors).

The most important of these are the light water reactors, as they are also operated in Germany at present. They have proved themselves worldwide, and are the reactors of choice not least because of their safety aspects. Apart from a few exceptions, light water reactors are the only ones that have been designed and built worldwide for some years now.

Under the Atomic Energy Act [33] (AtG) §7 — approving plant — para. 3, consent is required to decommission plant and safely contain the ultimately decommissioned plant or demolish plant or sections of plant.

Apart from the Atomic Energy Act, there are other statutory foundations and nuclear regulations to be considered:

— Radiation protection regulations (in German: StrSchV)

— Atomic Energy Act procedural regulations (in German: AtVfV)

— Law on environmental compatibility testing (in German: UVPG)

The ‘decommissioning guidelines’ (guidelines for decommissioning, safe containment and demolition of plant or plant components under §7 of the Atomic Energy Act) [34] are designed to bring the relevant aspects of approval and regulation together. It is also intended to create a common understanding between the Federal Government of Germany and Federal States on proper performance and harmonising existing views and methods.

There are also BMU guidelines, reactor safety committee (RSK) recommendations, Nuclear Safety Standards Committee rules (KTA), radiation protection committee (SSK) rules and relevant conventional rules to be taken into account when planning and implementing dismantling.

The rules to be observed when using anchor plates with headed studs are as follows:

— European Technical Approval for steel plates with cast-in headed studs for the manufacturer in question, e. g. ETA 03/0041 for Nelson headed studs [68]

— ETAG 001, Guidelines for European Technical Approval of metal anchors for use in concrete, Annexe C, Design methods for anchorages, February 2008 [65],

— DINENISO 13918:2008, Welding-Studs and ceramic ferrules for arc stud welding [73]

— DIN EN 14555:2006, Welding — Arc stud welding of metallic materials [74]

— DIN EN 10025-1:2005, Hot-rolled products of structural steels — Part 1: General technical delivery conditions [69]

— DIN EN 10088-1:2005, Stainless steels — Part 1: List of stainless steels [70]

— Deutsches Institut fur Bautechnik, using anchors in nuclear power plants and nuclear installations, guidelines for assessing anchor fastenings when granting approvals in individual cases under German Federal States Building Regulations 1998:09 [63]

— Deutsches Institut fur Bautechnik, Guidelines for anchor fastenings in nuclear power plants and other nuclear installations, June 2010 [67]

— DIN SPEC 1021-4-2:2009, Designing anchoring of fastenings for use in concrete — Part 4.2: Headed studs, German version CEN/TS1992-4-2:2009 [75].

3.2.1.1 General notes

Building structures in recent nuclear power plants are now expected to last for 60 years in operation, and even more than 80 years if we include commissioning and shutdown, so ensuring the materials characteristics required over such long periods makes choosing the right materials particularly important.

3.2.1.2 Concrete Normal weight concrete

The concrete strength grades normally used in nuclear installations in Germany are C30/37, and in exceptional cases C35/45, as in site concrete in particular cases. Concrete is normally mixed on site.

It was initially thought to make financial sense to use concrete in strength classes C55/67, because of its high strength, but this has proved to be less robust than expected, for various reasons. High-strength concrete is less ductile: any cracks which occur develop straight through the aggregates, creating relatively smooth crack surfaces. This affects integrity and ‘self-healing’ considerably.

For inland waterways, KTA 2207 [23] assumes a flood runoff with an exceedance probability of 10~4/a. This flood runoff can be determined either purely on a basis of probabilities or by extrapolating from statistics available. KTA 2207 uses this extra­polation, which is based on the Kleeberg and Schumann method [44]. This extrapolates from a peak level water runoff with an exceedance probability of 10~2/a to a peak level water runoff with an exceedance probability of 10~4/a.

This flood runoff value obtained, finally, gives the design basis water level from a corresponding water level runoff relationship for the location concerned.

As fastenings cannot be accessed once they are installed, safety-related components in particular are subject to particular corrosion protection requirements, depending on the ambient conditions involved.

Inner areas will very generally be dry, so under these conditions, protective coatings or galvanising provides sufficient corrosion protection.

Stainless steel materials must be used in outside areas and in inside areas where there are corrosion factors. Under DIN EN ISO 13918 [73], headed studs are in group SD3 with materials 1.4301 and 1.4303 to DIN EN 10088 [70]. For steel plates, materials 1.4571 and 1.4401 to DIN EN 10088 are used.

Where particularly corrosive influences are present, such as chemical pollution, stainless materials must be checked to see whether they can be used in each case.

Anchor plates made of currently standardised materials must not be used in chlorinated atmospheres.

Under the agreement between the Federal Government of the Federal Republic of Germany and the utility companies of June 2000 and the subsequent amendments to the Atomic Energy Act in April 2002, so-called on-site (decentralised) interim storage facilities were built at nuclear power plant sites between 2004 and 2007.

Decentralised interim storage facilities are those in which burned-out fuel elements are kept under controlled conditions at nuclear power plant sites for relatively long periods before being moved to final storage.

Interim storage facilities can be divided into two basic types:

— WTI design

Lightweight double-bay hall structures, walls approx. 70 cm thick, roof slabs approx. 55 cm thick, double-bay buildings consisting of two halls separated by a partition wall.

This model is based on the interim storage facilities at Gorleben, Ahaus and Lubmin/- Greifswald (northern interim storage facility).

Integrated operating areas with two cranes, stored in double rows (Figures 4.8 and 4.9)

— STEAG design

Solid single-aisle hall design with walls approx. 1.20 m thick, roof slabs approx. 1.30 m thick with separate operating building, one crane, compact storage (Figures 4.10 and 4.11).

The STEAG design was developed in view of using more cost-effective containment models in the future.

In accordance with the multiple barrier principle in nuclear technology the strength­ened building structure and future containment generation are designed to serve as additional barriers.

Both models share the same basic features: single-storey reinforced concrete halls with wall and roof slab openings for natural cooling. Inside the halls, a partition wall separates the reception/trans-shipment area from the storage area. Both storage designs have 1401 crane systems; the WTI halls need two of these because of their two-bay structure. For the building design of interim storage facilities see Section 4.3.2.3.

Once taken into store, the containers, which essentially contain irradiated fuel rods, can be described approximately as follows: height 6.50 m, diameter 2.80m and a dead weight of 1251. The container walls in the cylindrical and floor areas are approx. 420 mm thick.

The containers are sealed tightly using a cover system, using mainly CASTOR V/19 (Castor: cask of storage and transport of radioactive material) containers to date. These containers can hold up to 19 fuel elements (Figure 4.12).

The top of the container body is stepped to take the cover. At the head and foot of the container body are two overlapping carrying frames to which the storage hall crane lifting gear can be attached.

Containers are transported by rail or road exclusively and delivered to the interim store. Storage containers are transported horizontally to be stored in the interim store.

To unload containers, the storage hall crane attaches to them via the carrying frames provided, and the transporter vehicle takes them. Containers are then driven to their preset storage positions, set down upright and connected to a container monitoring system.

The current steelwork standards (DIN 18800-1 [55] and/or DIN EN 1993-1-1 [56]) with the new partial safety factor have only been reflected in KTA rules [14] for steel structures to a limited extent to date: so KTA status report KTA-GS-78 [51] advises relating steelwork load cases H, HZ, HS1, HS2 and HS3, and requirement categories A1, A2 and A3 (cf. Table 6.4).

How steelwork structures are designed depends on which KTA rule is to be used for the structure in question: so verifications may be required either by the global safety concept or the partial safety concept.

Fundamentally, the design procedures in DIN 18800-1 may be used (Table 6.5). Stability verifications must also be considered here, such that either with beam structures the limits of slenderness must be observed in all cross-sections, or with plates and shell structures buckling safety must be verified to DIN 18800-3 [57] or DIN 18800-4 [58].

The plastic-plastic design procedure, as shown in Table 6.5, reflects the plastic hinge analysis as a simplified method. More precise design procedures, such as using non­linear calculation methods reflecting realistic steel material laws, may also be used. When using plastic cross-section or system reserves, the design criteria in Table 6.5 must be observed.