Description of natural circulation core cooling system

The core is cooled mostly by the single-phase free convection due to natural circulation. The natural circulation loop consists of the core, the chimney, the downcomer and the lower plenum. In order to establish natural circulation at different RPV liquid levels during normal operations and accidents, the core barrel has large communication-holes above the SG top elevation and small holes at several elevations below that. The bypass flow through the small holes is approximately 5% of the core flow during the full-power operational condition. Since the RPV coolant is self-pressurized, that is, the pressure is determined by the temperature at the liquid level in the RPV, the core outlet is saturated. The fuel assembly is based on the 17×17 fuel assembly design with Zircaloy-4 cladding UO2 pellets that is used for the current PWRs except for the fuel pin pitch. The pitch of the PSRD is 13.9mm, which is 1.3mm wider than that of the current PWRs to enhance burn-up by increasing the moderation effect. This geometry decreases the flow resistance along the core.

TABLE XVin-1. MAJOR CHARACTERISTICS OF THE PSRD-100

Reactor power (MWt) Power output (MW(e))

100

27 to 31

Reactor coolant

Operation press. (MPa) Inlet/Outlet temp. Flow rate (kg/s)

10

270.4/311

450

Reactor core Diameter/height (m) U235 enrichment (Wt%) Fuel inventory (t)

1.62/1.50

~4.9

7.1

Fuel

Outer dia./pitch (mm) Burnable poison No. of fuel assembles

9.5/13.9

9%Gd2O3

37

Control rod and CRDM

Absorber

No. of control rods

B4C 24 x 37

Steam generator

Type

Temp./press. (°C/MPa)

Once-through

289/4.0

Reactor vessel Inner dia./height(m)

4/10

Containment

Type

Design press. (MPa) Inner dia./height(m)

Water-filled

2

7.3/14.6

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FIG. XVIII-1. Concept of PSRD

So far, the PSRD natural circulation core cooling has been evaluated for steady and transient states by using the thermal-hydraulic analysis codes. During such analyses, the load following capability was analyzed using the RETRAN-02 code by decreasing the feedwater flow rate from 100% to 50% in 200 sec, keeping it at 50% for 200 sec, and then increasing from 50% to 100% in 200 sec to represent a typical day-load change as shown in Fig. XVIII-2(a) (ISHIDA, T., 2003-2). The automatic reactor control system was not used to clarify the inherent self-controllability of the system. The calculation results showed that the reactor power responded well with the delay time of 50 to 100 sec and the overshoot of up to 10% as shown in Fig. XVIII-2 (b). The response of the natural circulation flow rate was also stable without showing high-frequency oscillations as shown in Fig. XVIII-2 (c). The results indicated the inherently stable nature of the PSRD natural circulation cooling system. The results also indicated that the responses will be more stable for the actual load follow condition with the operation of the reactor automatic control system without excessively depending on the use of the control rod operation.

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