Figure 2.3.1 9x9 Fuel assembly (a type) and fuel rod structure (example)

outer surface is determined by the bulk coolant temperature and the heat transfer coefficient between the cladding and the coolant. The heat transfer coefficient in the sub-cooled region (single phase convection region) differs greatly from that in the nucleate boiling region (Section 2.3.1-(2) item iii) a). Any drop in heat transfer coefficient due to CRUD formation and oxide film formation on the cladding surface should also be taken into account. Once the outer surface temperature of the cladding is known, its inner surface temperature can be obtained by the thermal conductivity of the Zircaloy-2 cladding. Then, the pellet surface temperature is calculated from the thermal conductivity of the gap between the pellet surface and the inner surface of the cladding. Finally, by the thermal conductivity of the UO₂ pellet, the temperature distribution in the pellet can be estimated. In this series of calculations, the temperature dependence of thermal conductivities of both the UO₂ pellets and the Zircaloy-2 cladding are taken into account

The linear power density (thermal power

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Upper tie plate

Channel fastener

Expansion spring

Spacer

Stan dard fuel rod

Water rod

Lower tie plate

Figure 2.3.2

Channel box

(not a part of fuel bundle)

Fuel cladding

Partial length fuel rod

Plenum spring

Pellet

Fuel assembly structure (9x9 fuel A type)

Channel box

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High burn-up 8x8 fuel

9x9 fuel (A type)

Q '■ Standard fuel rod

■ Partial length fuel rod © : Water rod

Figure 2.3.3 Fuel rod arrangement (example)

output per unit length of a fuel rod) of a BWR fuel element is limited to a specified value to prevent the fuel center from melting down.

For example, the maximum temperature at the UO₂ pellet center of a 9x9 fuel assembly is approximately 1550°C at its maximum linear power density (44 kW/m) during normal operations. The melting point of the UO₂ pellets is about 2800°C. The melting point decreases slightly by the addition of Gd₂O₃ and progression of fuel burn up. But the maximum temperature at the fuel center is still below the melting point of the fuel when these effects are taken into account. (The maximum linear power density is defined as the maximum local thermal power output per unit length of a fuel rod in the core.)

2. Release of gaseous fission products

As the fuel burn up increases, part of the gaseous fission products diffuse out of the pellets, resulting in internal pressure increases within the fuel rod. The release of gaseous fission products depends on the fuel burn up and pellet temperature. In the fuel design, an empirical release model, derived from experimental data, is used to estimate the amount of gas released from the pellets.

3. Pellet-clad mechanical interactions (PCIs)

Since the radial thermal expansion of the UO₂ pellets exceeds that of Zircaloy-2, the pellets may come in contact with the cladding, causing internal stress on the cladding. In addition to the thermal expansion, the swelling of the UO₂ pellets due to formation of fission products (largely fission gases), also contributes to the expansion. The pellet-clad mechanical interactions are analyzed by considering both thermal expansion and swelling of the pellets and the thermal expansion of the cladding.

2. Mechanical design

1. Materials

Key structural materials of a BWR fuel assembly are selected with due consideration to their exposure to heat, radiation and the hydrochemical environment Cladding tubes, spacers and channel boxes are made of Zircaloy and the top and bottom tie plates are made of stainless steel (type 304). Springs made of Inconel type X750 are used in all spacer grids. Experiences in the past

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the cumulative stresses never reach the design fatigue limit throughout the cladding lifetime.

5. Fretting corrosion

It has been recognized that two metal surfaces in contact may be subject to corrosion or abrasion (fretting corrosion) when they are subject to slight vibrations in a high pressure and high temperature aqueous environment

In order to prevent the fretting corrosion, spacer grids are used to suppress fuel rod vibrations and fix the space between the fuel elements. No fretting corrosions have occurred in the existing designs of BWR fuel assemblies.

6. Hydrogenation

A small amount of water in the cladding may interact with the Zircaloy cladding on its inner surface and form hydrides. This has been known to result in cladding failures. Therefore, the fuel element fabrication process is carefully controlled to keep the water contents low. Increased iron (Fe) content in the cladding material is being tested as a design measure against hydrogenation.

7. Bending of the fuel rod

During the fabrication process residual stresses in the fuel cladding and channel box are removed. Fuel rods are restrained horizontally by the spacer grids and the upper and lower tie plates allow axial elongation of the fuel rods due to thermal expansion or irradiation growth.

The spacer springs are designed so that their contact forces will not give axial constraint to the fuel rods and allow for their free axial elongation in order to prevent bending of the fuel rods.

8. Creep budding of cladding

The fuel cladding gradually becomes vulnerable to buckling (creep collapsing) due to external pressure and elevated temperature during the reactor operation. Therefore, the cladding is designed with small ovality tolerances to prevent the creep collapse. No creep collapse of the BWR fuel cladding has been experienced so far.

9. Local mechanical interaction between the pellet and the cladding

During the reactor operation, cylindrical UO₂ pellets deform to an hourglass shape (the diameters at the top and bottom ends get larger than at the center), causing local strains on the cladding at their contact spots with the pellet

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edges. This may eventually lead to cladding failures. To minimize the local mechanical interactions between the pellets and the cladding, edge chamfered pellets and heat treated cladding materials with enhanced ductility due to recrystallization and annealing are in use.

j. Handling and transportation of the fuel assembly All the components composing a fuel assembly including the upper and lower tie plates, the cladding and the spacer grids are designed to have sufficient strengths to withstand the loads expected during the fuel handling and their transportation.

3. Thermal-hydraulic design

1. Transient criteria

Nuclear fuel must maintain its integrity not only during the normal operation, but also during abnormal transients (such as a reactor over­power or insufficient fuel cooling) caused by a component failure or an operational error of the reactor system. The design criteria to be met by the fuel elements during the abnormal transients are called the transient criteria, and they are determined taking into account the following two types of fuel failure mechanisms:

1. cladding failures caused by over-heating due to insufficient cooling (cooling-failure transients); and

2. cladding failures caused by the permanent strains induced by expansion differences of the UO₂ pellets and the cladding (FCIs).

The first type of cladding failures (a) can be avoided by preventing the transition of boiling heat transfer mechanism from the fuel to the coolant (i.e., transition from nucleate boiling with high heat transfer coefficients to film boiling with low heat transfer coefficients). The heat flux which initiates the boiling transition in a BWR fuel assembly is called the critical power and the critical power divided by the actual operating power (heat flux) of a fuel assembly is the critical power ratio (if the critical power ratio is larger than one, the operating power of the fuel assembly concerned is below its critical power). The lowest critical power ratio among the fuel assemblies is called the minimum critical power ratio (MCPR). The minimum critical power ratio is not 1.0, but slightly larger than 1.0. The value

is set, taking into account the core monitoring parameters measured during plant operations with statistical analyses of their errors, so that the fraction of fuel rods with no boiling transition is less than 99.9%. Hie design MCPR in the current BWR design is set at about 1.07.

The Japan Atomic Energy Society published a methodology to evaluate the fuel integrity under transient boiling transitions, “Fuel integrity evaluation criteria in the wake of transient boiling transitions in BWR: 2003.” The criteria require that the cladding temperatures and the dry­out duration (the time duration after the boiling transition) must be demonstrated by means of event analyses as not exceeding the pre-specified limitations, when evaluating fuel integrities at the converged state of the event or in handling the fuel assemblies after the event Another possible unusual event, which may cause this type of cladding failure, is a reactivity anomaly event inducing rapid load changes (nuclear excursions). The safety limitations are defined as the fuel heat rate (the amount of heat generated by the fuel) for preventing the cladding failure. Further elaboration is dispensed with here, since this is not directly related to the fuel design.

The second type cladding failures (b) assumes that the difference in radial expansion between the fuel pellets and the cladding induces excess strain in the cladding and possible eventual failure. A radial plastic strain of one percent (1.0 %) is set as the transient criterion to be met by the fuel cladding design to prevent this type of cladding failure. This criterion has been derived from the analyses of the break down test results of the internally pressurized Zircaloy cladding at elevated temperatures. The second type cladding failures can be prevented by limiting the maximum power (heat rate of the fuel) during transients, because the difference in expansion is mainly governed by the heat rate of the fuel,

2. Fuel design criteria under normal operation

Design criteria are set for fuel under the rated power operation condition of the reactor, in order to prevent the fuel from violating the above mentioned transient criteria, in case any abnormal transients are caused by a component failure, operational error of the reactor, or other reasons.

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