Fuel Rod Analysis

[1] Fuel rod integrity

The principle role of fuel rod cladding is to confine radioactive FPs and to prevent contamination of the coolant system; therefore an assurance of fuel rod integrity is important for reactor design and safety.

When an abnormal transient event occurs in a nuclear plant with a frequency of more than once during the plant life time, the safety criterion is associated with whether the core can return to the normal operation without core damage

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image340Suppression Pool

Fig. 2.48 Calculation model for LOCA reflooding analysis

after the plant is stabilized. Thus, fuel rods are required to keep their integrity during abnormal transients (anticipated operational occurrences) as well as normal operation.

The criteria for fuel rod integrity differ somewhat between BWRs and PWRs because of different operating conditions. Fuel rods are designed generally based on the following criteria [25, 27].

(i) Average circumferential plastic deformation of fuel cladding < 1 %

(ii) Fuel centerline temperature < Pellet melting point (PWR)

(iii) No overpressure within fuel rod

(iv) Allowable stress of fuel rod cladding

(v) Cumulative damage fraction < 1

The overall fuel rod integrity is evaluated further in cladding corrosion and hydriding, pellet cladding interaction (PCI), cladding creep rupture, rod fretting wear, cladding bending, and irradiation growth of the fuel rod and assembly.

The fuel damage criteria and allowable limits of BWRs and PWRs are given in Table 2.2. Based on the system of fuel damage occurrence in BWRs or PWRs, restrictions at normal operation and allowable limits at abnormal tran­sient are determined in the fuel design considering the each following;

(i) Cladding damage due to overheating resulting from insufficient cooling

(ii) Cladding damage due to deformation resulting from a relative expansion between pellet and cladding

Table 2.2 Criteria and allowable limits of fuel damage in BWR and PWR

Damage

BWR

PWR

(i) Cladding

Criteria

The fuel cladding integrity

The DNB design basis

overheating

ensures that during normal operation and abnormal transients, at least 99.9 % of the fuel rods in the core do not experience transi­tion boiling

requires at least a 95 % probability, at a 95 % confidence level, that the limiting fuel rods in the core will not experience DNB during normal operation or any transient conditions

Allowable limits at transient

MCPR: 1.07

MDNBR: 1.30

Restrictions at normal

MCPR: 1.2—1.3

MDNBR: 1.72

(ii) Cladding

Criteria

The fuel cladding integrity

The fuel cladding integrity

deformation

ensures that during normal operation and abnormal transients, the fuel rods in the core do not experience circumferential plastic deformation over 1 % by PCI

ensures that during normal operation and abnormal transients, the fuel rods in the core do not experience circumferential deforma­tion (elastic, plastic, and creep) over 1 % by PCI

Allowable

Direct evaluation of criteria

MLHGR: 59.1 kW/m

limits at transient

by experiments

(2,300 °C fuel centerline temperature)

Restrictions at normal

MLHGR: 44 kW/m

MLHGR: 43.1 kW/m

(1,870 °C fuel centerline temperature)

For damage mechanism (i), the BWR criterion is that “the fuel cladding integrity ensures that during normal operation and abnormal transients, at least 99.9 % of the fuel rods in the core do not experience transition boiling”. In PWRs, similarly, the criterion is that “the DNB design basis requires at least a 95 % probability, at a 95 % confidence level, that the limiting fuel rods in the core will not experience DNB during normal operation or any transient conditions”. (DNB means the departure from nucleate boiling.)

Based on those criteria, the BWR design requires the allowable limit of the minimum critical power ratio (MCPR) to be 1.07 at abnormal transients and the restriction to be 1.2—1.3 at normal operation. Similarly, the PWR design has the allowable limit of minimum heat flux ratio (minimum departure from nucleate boiling ratio, MDNBR) to be 1.30 at abnormal transients and the restriction to be 1.72 at normal operation.

For the damage mechanism (ii), the BWR criterion is that “During normal operation and abnormal transients, the fuel rods in the core do not experience circumferential plastic deformation over 1 % by PCI”. In PWRs, similarly, the criterion is that “During normal operation and abnormal transients, the fuel rods

Table 2.3 Fuel rod behavior in FEMAXI-6

Thermal behavior (temperature dependent)

Mechanical behavior

Pellet

Thermal conduction (radial heat flux distribution)

Fission gas release (temperature and burnup dependent)

Thermal expansion, elasticity and plasticity, creep, cracking, initial relocation, densification, swelling (solid FPs and gas bubbles), hot-press

Cladding

Thermal conduction Oxidation film growth

Thermal expansion, elasticity and plasticity, creep, irradiation growth, oxidation film growth

Fuel Rod

Gap thermal conduction (mixed gas, contact, radiation) Cladding surface heat transfer Gap gas flow

PCMI (pellet cladding mechanical interaction), friction, bonding

in the core do not experience circumferential deformation (elastic, plastic, and creep) over 1 % by PCI”.

Central melting of pellets in LWR fuel causes a phase change and a volume increase, and the fuel cladding may be substantially deformed mainly due to pellet cladding mechanical interaction (PCMI). The fuel rod design criteria of PWRs require that the fuel centerline temperature will be lower than the pellet melting point. Hence, the allowable limit of the fuel centerline temperature is determined to be 2,300 °C at abnormal transients and its corresponding max­imum linear power density is 59.1 kW/m. The restrictions at normal operation are 1,870 °C for the fuel centerline temperature and 43.1 kW/m of the maxi­mum linear power density.

For internal pressure of fuel rods, the BWR design requires that the cladding stress due to the internal pressure will be less than the allowable strength limit. The PWR design restricts the internal pressure of fuel rods to less than the rated pressure of primary coolant (157 kg/cm2-g) in order to avoid expansion of the gap between pellet and cladding due to creep deformation of cladding outside at normal operation. This phenomenon is called “lift-off”. The criteria of “ASME B&PV Code Sec III” are used as allowable stress limits for LWRs.

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