Criteria to ensure thermal integrity of the fuel

PHWR fuel can potentially operate at high linear heat generation rates, in comparison with other reactors such as LWRs. Moreover, the heat generation rate can increase even further locally due to effects such as end flux peaking. For fuel pellets located near fuel bundle junction areas (where bundles are in contact with adjacent bundles), the fission rate increase may be as high as 10 to 15%. The increased heat generation, however, is balanced by the increased heat

transfer by axial as well as radial heat flow from the end pellet. These conditions can lead to increased peak pellet temperatures, putting end pellets at greater risk of melting during an event.

Reported criteria

(a) During normal operations and during AOOs: The maximum temperature has to be less than the melting temperature of UO2 for 90% of the pellet cross-section at the hot spot. Fuel melting temperature limit is approximated as 3073 K (2800ºC). The decrease of the melting point with burnup is neglected in the calculations (for vertical channel-type PHWRs, Argentina);

(b) The centreline pellet temperature has to be less than stoichiometric UO2 melting temperature (e.g. 3120 K ± 30 K [26]) (Canada);

(c) The centreline temperature has to be below the melting temperature. The latter is given by:

T = 2805  0.133 × ω (1)

where T is the melting temperature in °C and ω is the fuel burnup in MW·h/kgU;

this accounts for experimental uncertainties and burnup effects (China);

(d) During normal operations and during AOOs: The calculated maximum fuel pellet centreline temperature, with due allowance for irradiation, tolerances, uncertainties, etc., has to remain below the melting point (India);

(e) During normal operations and during AOOs: Centreline melting is used for fuel melting indication. It is evaluated for both ends of fuel rods at the outer ring which experience the highest neutron flux. Pellet melting temperature is 3078 K (2805°C) for fresh fuel and decrease at the rate of 32 K per 10 MW·d/kgU. Margin needs to exceed 20%. This evaluation needs to done at both ends of fuel rods in the outer ring which experience the highest neutron flux (Korea);

(f) The fuel centreline temperature needs to be much below the melting point of UO2

which is 3033 K (2760^oC) (Pakistan);

(g) Criterion is based on centreline melting. Actual melting occurs at 3113 K (2840^oC);

for conservatism, 3038.15 K (2765^oC) is used for a burnup of 300 MW·h/kgU (Romania).

Discussion

The value of 3120 K ± 30 K [26] for the melting point of stoichiometric UO2 is the value recommended by Rand et. Al. [27] from their analysis of fourteen experimental studies (over a period of 20 years) and has been accepted internationally. The value of 3120 K ± 30 K is used by MATPRO [28] as input in the calculation of the melting temperature of an uranium dioxide pellet as a function of fuel burnup.

For normal operations, pellet centreline melting is not permitted. If the pellet melts, the resulting volumetric expansion of the pellet may potentially push the sheath past breaking. Further, if the molten UO2 flows, it can potentially reach the Zircaloy and melt it too.

For normal operations, the maximum pellet centreline temperature is usually considered to be below the pellet melting point with adequate margin which typically bounds the uncertainty of the calculation correlations.

3.1.2.2. Overheating of the fuel sheath: critical heat flux

Overheating of the fuel sheath can quickly degrade the strength of the sheath material. It can also lead to rapid oxidation, crevice corrosion, overstrain, creep and in worse case melting of the fuel sheath. Excessive bowing can occur if the overheating is localized and large. Excessive local bowing can potentially damage a neighbouring fuel element and even the pressure tube.

While liquid dryout on fuel sheath surface does not necessarily cause an abrupt rise in fuel surface temperature, a prolonged dryout may lead to a level of overheating that triggers the damage mechanisms mentioned above, resulting in the fuel (element/bundle) no longer meeting some of its design requirements, and therefore no longer being fit for service.

Reactor core design and operation apply limits on some of the thermohydraulic parameters to ensure adequate cooling to the fuel. In particular, the thermohydraulic design of the fuel assures that specified thermal safety limits are not exceeded in operational states. Although these thermal limits are not bounds on parameters directly related to properties describing the damage mechanisms mentioned above, their values are nevertheless fuel bundle design dependent, and their formulation are among the defence in depth provisions (measures) taken to address the damage mechanisms mentioned earlier, and ensure that the safety objectives formulated in Section 2.2.1.1 are met; for those reasons, these thermal limits are reported as fuel acceptance criteria for operational states.

A thermal safety limit is commonly specified as a margin to the critical heat flux (CHF), such as a minimum CHF ratio. The CHF is the thermal limit, beyond which heat transfer from the heated surface to the coolant reduces significantly, causing a rapid rise in surface temperature.

The CHF ratio is the ratio of CHF over local heat flux. In general, empirical methods, such as empirical correlations, are employed to predict CHF values. These empirical CHF methods are bundle-design specific and are derived on the basis of fuel bundle simulation experiments, covering the range of thermohydraulic conditions (e.g. coolant pressure, temperature, flow) encountered in all the plant states (normal operations and accident conditions) considered in the design of the nuclear power plant. Sufficient margin to CHF is preserved to address uncertainties of various parameters affecting CHF or CHF ratio.

Reported criteria

(a) The departure of nucleate boiling ratio has to be 1.25 minimum. There are additional margins applied to account for the uncertainty in the initial normal operation conditions and also in case of AOO coolant pumps shut-off (Argentina);

(b) Under normal operating conditions, the fuel bundle has to maintain sufficient margin to dryout. No quantitative thermal safety limit is specified in terms of avoiding occurrence of dryout on fuel sheath surface. However, a sufficient margin to CHF is demonstrated to accommodate various uncertainties. In Canada, avoiding fuel sheath dryout, under normal operating conditions, is a fuel bundle design requirement (Canada);

(c) Default CATHENA heat transfer and critical heat flux (Groeneveld lookup tables) correlations are used. For post-dryout, the Groeneveld-Delorme correlation is used (China);

(d) The minimum CHF ratio has to be 1.3 under normal condition and 1.1 under anticipated operational occurrences (India);

(e) The critical power ratio has to be ≥1.0 and Xc-BL correlation is used (Korea);

(f) The maximum heat flux in the hot channel is limited to 1.08 MW/m² (108 W/cm²) and the corresponding clad temperature of 322°C at normal coolant flow (Pakistan);

(g) The peak sheath temperature has to be less than 350^oC for normal operating conditions, with sufficient margin to the limit for onset of sheath dryout using Groeneveld critical heat flux lookup tables. Sheath oxidation is simulated using Urbanic-Heidrick correlations (Romania).

Discussion

Water cooled nuclear reactors (with the exception of supercritical pressure reactors) are limited in power to avoid occurrence of CHF under normal operations, AOOs and some event conditions. Critical heat flux is the thermal limit of a heated surface, reached due to phase change of liquid coolant, such as bubble formation, bubble coalescence, and vapour film formation, which abruptly decreases the heat transfer, causing localised overheating of the surface. The behaviour of CHF depends on flow conditions. Under subcooled or low quality conditions, such as those encountered in PWRs, CHF occurs at a relatively high heat flux and appears to be associated with the cloud of bubbles adjacent to the heated surface which reduces the amount of incoming water and acts like an insolation layer on the surface. When this behaviour occurs, surface temperature abruptly rises to a high value. This kind of CHF can be classified as departure from nucleate boiling, commonly encountered in PWRs. Under high quality conditions such as those in CANDU-type PHWRs, CHF occurs at a lower heat flux.

The flow pattern prior to CHF is usually annular, and the fuel surface is normally covered by a liquid layer. When the evaporation rate is high enough, the liquid layer can no longer be sustained and dry patches will develop on the fuel surface, which reduces heat transfer from the surface to the coolant. Since the velocity in the vapour core is high, post-dryout heat transfer is much better than for low quality cases; wall temperature rises are moderate and less rapid. In general, the dryout-type CHF is of interest to PHWRs.

As mentioned earlier, a prolonged dryout may trigger various fuel sheath damage mechanisms including melting of the fuel sheath. For this reason, prevention of fuel sheath dryout (or dryout-type CHF) has been chosen by many States as a thermal safety limit for normal operation, some AOOs and some DBAs of horizontal channel-type PHWRs. In other scenarios where exceedance of CHF is permitted, post-CHF heat transfer calculations are performed to evaluate the fuel sheath temperatures (the influencing parameter for some other fuel criteria).

The thermohydraulic design of horizontal channel-type PHWRs is limited by the necessity to maintain an adequate safety margin between the operating heat flux and the CHF of a fuel bundle. Given that neither the operating heat flux nor the CHF is measurable in the reactor core, the critical channel power has been used as a surrogate parameter of CHF in maintaining the necessary margin to CHF. The critical channel power is the channel power at which the fuel

string¹⁶ starts experiencing fuel sheath dryout anywhere in the fuel string. It is simulated using a computer code, based on the knowledge of CHF, the axial power profile, and the thermohydraulic conditions of the reactor core.

In operation of horizontal channel-type PHWRs, channel powers are related to the reactor power and are measurable. Therefore, the necessary safety margin between the operating heat flux and the CHF of a fuel bundle is ensured by maintaining an equivalent margin between the channel power and the corresponding critical channel power, or by maintaining a minimum critical power ratio (the ratio of critical channel power over channel power). The minimum critical power ratio encompasses the margin of operating heat flux to CHF of fuel bundles. No information on this criterion was specifically requested from the PHWR States.

As discussed in Appendix II, which covers the Canadian approach to the formulation of fuel fitness-for-service criteria, CHF may be reached for a very short period of time, and for low sheath temperature (<450ºC), without causing damage (‘not damaged’ as defined in Section 2.2.1.1) to the fuel; this fact has been used to derive the fitness-for-service criteria for AOOs presented in that Appendix.

A summary of the current state of knowledge on fuel thermal hydraulics for CANDUs can be found in [29, 30].

3.1.3. Criteria to ensure structural integrity of the fuel