Civil Engineering Reference
In-Depth Information
5.6 Detailed Design Requirements at Safety Level 1
At safety level 1 (Sect.
5.3.2.1
), the key role is played by the thermodynamics,
neutron physics and mechanical design of the nuclear reactor and the properties of
used materials in components, such as the reactor pressure vessel, pumps and pipes.
In addition, training of the operating personnel must be ensured.
5.6.1 Thermodynamic Design of LWRs
For achieving higher redundancy, the plant design is split into several identical
coolant systems for heat removal from the reactor core. Present pressurized and
boiling water reactors have three or four identical cooling circuits with associated
coolant pumps, steam generators, feedwater systems, emergency core cooling
systems etc. connected to the reactor pressure vessel. Figure
5.3
shows the primary
coolant circuit system of a modern PWR as described in Chap.
3
. This includes the
pressurizer for coolant pressure control and stabilization. In the pressurizer, electric
heaters increase pressure through evaporation of the pressurized water. A
pressurized-water spray system in the pressurizer condenses the steam, thereby
lowering the pressure. When the pressure becomes too high, relief valves above the
pressurizer can automatically release steam into an expansion vessel in the reactor
containment and thus prevent overpressure failure of the primary cooling system.
The coolant pressure in the cooling circuits and in the pressure vessel of a PWR
is chosen such (15.5 MPa) that nominal power of the fuel rods of the reactor core
cannot give rise to local or subcooled boiling. In addition, there must be a sufficient
margin relative to the critical heat flux. Because of corrosion and embrittlement
problems, the temperature of the zircaloy cladding of a fuel rod should not exceed
350
C[
10
-
12
].
Critical heat flux on the surface of a fuel rod would give rise to departure from
nucleate boiling (DNB). At this critical level of the heat flux a vapor film is
produced on the surface of the fuel rod. This causes the temperature on the surface
of the fuel rod to rise so strongly as to cause failure (break) of the zircaloy cladding.
The Departure from Nucleate-Boiling Ratio (DNBR) is defined as the ratio between
the critical heat flux and the current heat flux on the surface of the fuel rod:
q
}
ð
critical
Þ
DNBR
¼
q
}
ð
actual
Þ
where q" is the heat flux [W/cm
2
] on the surface of the fuel rod.
In the design of the PWR core, this ratio is chosen as DBNR
¼
1.80. The critical
heat flux is determined on the basis of empirical relations [
10
,
11
].
As a design criterion for the maximum power of a fuel rod, the associated central
fuel rod temperature must not reach the melting point of UO
2
fuel (2,865
C) under
