so that, a permanent contact between the grid springs and the fuel rods over the life of the fuel assemblies is maintained. In addition, to avoid fuel rod contact with top and bottom nozzles (due to fuel rod elongation during irradiation), sufficient end clearances between rods and nozzles are provided.

Hold down forces of sufficient magnitude to oppose the dynamic lift of the primary coolant flow on the fuel assemblies are obtained by the leaf springs of the top nozzle. Besides the leaf spring forces, loading imposed by rapid insertion of the control rods (scram) as well as the hydraulic forces developed in the event of a postulated LOCA are considered in design of the control rod guide thimbles. The maximum impact force on the support grids for the design-basis earthquake occurrence is calculated by a group vibration analyses method and the results are compared with the experimental results to ensure that the grid deformation will never disturb the insertion of the control rods into the guide thimbles.

2. Reactor and Reactor Core

As shown in Figure 3.3.4, the structural elements of a PWR and the reactor core consist of the RV,

fuel assemblies, core internals (CI), rod cluster control (RCC) and control rod drive mechanism (CRDM).

The initially loaded core is divided into regions of three different enrichments. Fuel assemblies are arranged in a roughly cylindrical pattern in the core. Light water is used as the primary coolant to cool the core and moderate the neutrons at the same time. Boric acid, a soluble neutron absorber, is used in the coolant to control long term reactivity changes.

1. Structure of reactor and reactor core

The reactor internals are designed to support the fuel assemblies in the core and they are divided into two main parts, i.e., the lower core structure and the upper core structure. The primary coolant enters into the RV through the inlet nozzles located in the upper section of the vessel, passes down through an annular passage between the core barrel and the vessel wall (downcomer) to the lower plenum formed by the bottom head, then enters the bottom of the core and flows up to the upper plenum. The coolant temperature rises by absorbing the heat generated in fuel rods while passing through the

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Core barret

Upper core support plate

Control rod drive mechanism

Reactor vessel head

Upper core plate

Thermal shield

Core baffle

Reactor vessel

Fuel assembly

Radial support

Lower core support plate

Reactor vessel

inlet nozzle

Lower core

support plate

Lower core plate

Reactor vessel outlet nozzle

Upper core support column

Control rod cluster (withdrawn)

Figure 3.3.4 Reactor and internal structures core. After mixing in the upper plenum, the coolant leaves the RV through the outlet nozzles located on the same plane as the inlet nozzles.

The upper core support structure, shown in Figure 3.3.5, consists of the upper core support plate, upper core plate, rod cluster control (RCC)

Control rod cluster

guide Lube

guide tubes and thermocouples (for measuring the coolant temperature). Hie upper core support structure is positioned on the core barrel of the lower core support structure by flat-sided pins pressed into the core barrel. The major support structure of the reactor core is the lower core support structure shown in Figure 3.3.6. The core barrel of the lower core support structure provides an annular passageway for the coolant flow.

In order to minimize the corrosion products and accompanying radioactivity accumulation in the primary coolant, all parts which are in direct contact with the reactor coolant are made of either stainless steel or stainless steel-clad materials. The fuel cladding is made from Zircaloy-4 or zirconium- based alloy which has excellent corrosion resistance in addition to good nuclear characteristics. Water quality in the closed cycle of the primary cooling system is maintained to minimize corrosion. This is accomplished by adjusting the pH of the water to a proper value using lithium hydroxide (LiOH) and also, by supplying hydrogen gas to the water to promote recombination of the free oxygen molecules produced by dissociation of water molecules due to irradiation.

Figure 3.3.5 Upper core support structure

Figure 3.3.6 Lower core support structure

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Figure 3.3.7 Flow chart of core size determination

reactor core is standardized due to manufacturing reasons.

2. Nuclear design requirements for safety assurance

In order to ensure that the core has inherent safety provisions through the provided negative reactivity coefficient, and the plant has a sufficient shutdown margin, the following items are evaluated:

• Sufficient shutdown margin is provided by two independent reactivity control systems, i.e. by the RCC assemblies and by injection of boric acid.

• The design provides that the maximum reactivity

insertion rate corresponding to the maximum withdrawal rate of rod cluster control (RCC) never exceeds its upper limit.

• The design must guarantee a negative reactivity

feedback (Doppler coefficient and moderator temperature coefficient) characteristics of the core.

• Further, the reactor core is designed so that the core power distribution (power peaking coefficient) meets the criteria on fuel temperature and critical heat flux ratio (DNBR). After completion of the plant (i.e. completion of

the test run and start of commercial operation),

refueling is conducted during the periodical inspection outages, for which it is necessary to determine the next fuel loading pattern, recalculate the core nuclear characteristics parameters and confirm the safety of the core.

During reactor operation, it is necessary to trace the fuel burnup through the recording and analyzing of various data (specially, core output and core power distribution records). Fuel burnup information is used as input data to determine the fuel loading pattern in the next loading of the core.

The initially loaded core at the plant completion (first cycle core) and the partially refueled core after each refueling operation (second cycle, third cycle,...) are generally called the initial core and reload cores, respectively.

1. The initial core

The initial core is divided into three regions; each region is loaded with fuel assemblies of different enrichment on the premise of three- region-three-cycle refueling. The numbers of fuel assemblies in each region are almost the same. The fuel enrichment at each region is determined by considering the following items. CD Planned unloading fuel burnup.

2. Larger operation cycles higher fuel enrichment

3. Sufficient flattening of power distribution.

Fuel assemblies with the highest enrichments (region 3) are placed around the outside of the core (in a concentric circle) and the two groups of assemblies of lower enrichments (regions 2 and 3) are arranged in a selected pattern (checkerboard pattern) in the central parts of the core. The initial core is subject to three refueling operations which take place generally in accordance with an inward loading schedule. Because the initial core cycle contains more excess reactivity than subsequent fuel cycles as a result of the loading of all-fresh fuel, burnable poison rods are introduced to keep the negative moderator temperature coefficient (this is done by reducing the soluble poison concentration at the initial core cycle).

2. The reload cores

Reload core design calculations are carried out for each reload core after the second cycle core, since the design conditions listed below are different for different plants and different core

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Temperature Difference: Tw-Ts • “C

'l\v: Heat Transfer Surface Temperature Ts: Saturation Temperature

Figure 3.3.8 Boiling characteristics

corresponding heat flux is called the critical heat flux (or DNB heat flux). The critical heat flux not only depends on local conditions such as coolant flow, pressure and vapor content, but it also depends on many other factors such as inlet enthalpy of coolant, channel length, upstream channel shape and condition. For assessing the thermal margin of the fuel in the core, the ratio of the critical heat flux at a particular core location to the existing heat flux at the same location (the critical heat flux ratio: DNBR) is defined as nxmn critical heat flux local heat flux

DNBR is chosen as an index of the thermal margin of the fuels in the core.

2. The fuel rods are designed such that the fuel center temperature is less than the melting point of UO₂. This criterion is necessary to prevent the occurrence of the following problems.

(T) Sudden expansion of the uranium oxide fuel followed by pellet-clad interaction.

(2) Increase of the internal pressure of the cladding due to release of fission gases.

©Chemical interaction between fuel and cladding material.

Since the thermal conductivity of the UO₂ fuel is

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