Mitsubishi Heavy Industries Technical Review Vol. 49 No. 1 (March 2012)34

A New Concept for Critical Heat Flux Correlationsfor Safety Analysis of Pressurized-water Reactors

R e a c t o r C o r e E n g i n e e r i n g D e pa r t m e n t
Nuclear Energy Systems Headquarters

The most important priority of nuclear power plant design is “safety.” The accident at
Fukushima Dai-ichi Nuclear Power Plant following the Great East Japan Earthquake has
re-emphasized the importance of nuclear safety, requiring us to make a strong effort to improvesafety. Mitsubishi Heavy Industries, Ltd. (MHI) has been working on improving the accuracy andreliability of safety design evaluations to ensure safer and more flexible operations ofpressurized-water reactors (PWRs). In this context, MHI has developed and utilized a new concept
for critical heat flux correlations.

|1. Critical heat flux
A pressurized-water reactor (PWR) fuel consists of many fuel rods containing sintereduranium dioxide pellets sealed in zirconium alloy cladding. The fuel cladding provides animportant barrier to prevent fission products in a fuel rod from being released to the outside. The
fuel is cooled by high-pressure coolant (light water) at a temperature sufficiently lower than theboiling point. Therefore, it is unlikely that the fuel cladding will fail due to overheating unless asevere accident, such as a loss-of-coolant accident, occurs and results in core exposure. However,even with a sufficient amount of coolant in the reactor core, if a significant increase of local poweroccurs, a steam film can be formed by rapid steam generation on the fuel surface, which may cause
fuel failure due to overheating. Such a phenomenon is known as “departure from nucleate boiling(DNB).” One of the most important objectives of PWR safety analysis is to prevent DNB.
Critical heat flux (CHF) is the local thermal power where DNB occurs. The value of the CHF
depends on the pressure, temperature (i.e., enthalpy), flow velocity of the coolant and fuelspecification, especially the geometry and layout of the grid spacers, which hold the fuel rods in asquare array. For the safety design of the nuclear plant, CHF is measured against various coolant
conditions using the electrically heated test equipment that precisely simulates the actual fuel.
Based on a database obtained from the test, empirical DNB correlations are derived and are utilized
to predict CHF in the safety analysis.

|2. Background and characteristics of new correlations
Mitsubishi Heavy Industries, Ltd. (MHI) has been working on improving the analysis
technology for the core thermal-hydraulic behavior including DNB correlations. The enhancementof its accuracy revealed the margins involved in DNB predictions, and this has contributed tosafely increasing the core operating flexibility, such as extensions of fuel burnup and/or theoperating cycle length, as well as to improving the fuel economy. However, it also required thesafety analysis method to accommodate a wider range of applications.
Safety analysis technologies have been improved based on those originally developed in theU.S. in 1970s and 1980 s. MHI is replacing these conventional technologies including DNB
correlations with independently developed expertise with higher level applicability based on theanalysis experience and knowledge accumulated in MHI. These efforts are expected to contributeto the further development of safety analysis technologies.

Mitsubishi Heavy Industries Technical Review Vol. 49 No. 1 (March 2012)
35(1)Structure of the new correlation system
The new correlation system developed by MHI consists of multiple correlations
according to the fuel type and applicable conditions. In this stage, MHI has completed thedevelopment of the two basic correlations, the Mitsubishi generalized correlation for a standardgrid (MG-S) and the Mitsubishi generalized correlation for a non-vane grid (MG-NV), whilefurther development of the new correlations continues and will enhance the applicability of thesystem.
MG-S is the correlation applicable to MHI’s standard design fuel. The allowable range ofreactor operating conditions is defined using this correlation, and in the case of an abnormal
event, the reactor will be shut down by the safety-protection system before it exceeds thispredetermined operating limit. This correlation is also used in the safety analysis to confirm thevalidity of such safety systems.
The MG-NV correlation is used in the safety analysis of the special events, in which thereactor inadvertently restarts under a low-pressure and low-flow-rate condition after shutdown.
The MG-NV was developed based on the low-pressure and low-flow-rate data, to which theMG-S correlation is not applicable. Since the data simulating PWR fuel under low-pressure andlow-flow-rate conditions are limited, a conservative correlation was developed by referring to
the test data for the fuel geometries using grid spacers without mixing vanes, in which DNB is
more likely to occur than standard fuel using grid spacers with mixing vanes.
The new correlation set covers the entire range of coolant conditions assumed in thepresent PWR safety analyses. In addition, it provides reliable analysis even for low-pressure
and low-flow-rate conditions, where only a highly conservative evaluation was applicable since
the conventional correlations are not (Figure 1).

Figure 1 Applicable range of CHF prediction correlations

The current applicable range of the MG-S correlation is the same as that of the MIRC-1 correlation. The
applicable range of the MG-NV correlation covers the region corresponding to the conventional W-3 correlation,
while excluding unnecessary high pressures and high flow rates and including additional low pressures and low
flow rates.

(2)

Characteristics of the new correlations
The CHF generally decreases as the coolant enthalpy increases. Most of the existingcorrelations are based on a simple linear relationship between CHF and coolant thermal
equilibrium quality (i.e., the non-dimensional expression of enthalpy) due to the insufficient
data analysis technique at their origination in the 1970 s and 1980 s. Therefore, the applicabilityof these correlations is limited to a narrow range of thermal equilibrium quality. In creating theMG-S and MG-NV correlations, MHI analyzed the dependency of the CHF on the coolant
conditions in detail, and developed a technique to simulate the CHF characteristics against
thermal equilibrium quality with an appropriate curve (Figure 2).
Accordingly, the MG-S correlation is expected to be suitable for extending the databasein the future. The MG-S correlation can flexibly accommodate fuel design improvements thatare intended to enhance fuel performance to prevent DNB. Besides the flexibility for extendingMitsubishi Heavy Industries Technical Review Vol. 49 No. 1 (March 2012)36

the database range, the MG-S correlation has a high degree of accuracy (i.e., 17 % uncertainty),
equivalent to conventional correlations (Figure 3). In addition，the MG-NV correlation covers
low-flow-rate conditions, for which conventional correlations are not applicable due to theirlimited thermal equilibrium quality range. The MG-NV correlation has significantly improved
the CHF prediction accuracy for low-pressure and low-flow-rate conditions (uncertainty: 30%=> 19 %) (Figure 4). These two new correlations are expected to make large contributions
towards improving the reliability of the safety analysis.

Figure 2

Characteristics of new DNB correlations

Conventional DNB correlations predict CHF only for the limited thermal
equilibrium quality range. The new DNB correlations have been developed basedon a flexible curve fitting technology to represent the CHF behavior against
thermal equilibrium quality in a more appropriate manner.

Figure 3 Predictions
correlation

bytheMG-S

The predictions by the MG-S correlation agree wellwith test results. The uncertainties remain unchanged
from those of conventional correlations.

Figure 4 Predictions by the MG-NV
correlation
The predictions by the MG-S correlation agree wellwith test results including low-pressure and lowflow-rate conditions. Uncertainties are significantly
reduced in comparison with conventional correlations.

|3. Future prospects
MHI is continuously conducting research and development to enhance the safety and thereliability of nuclear power plants through improved safety analysis. The CHF data acquired in thisresearch will be incorporated into the database of the DNB correlation system. It is expected thatthe further extended database will result in a more reliable correlation system with a wider range ofapplicability.