BWR Core Design and Core and Fuel Management

3.2.1 General Core Design

[1] Features of BWR core [3]

BWRs operate in a direct cycle in which the steam formed in the core goes directly to turbines to generate electricity. A big difference from other type reactor cores is that there is a distribution of moderator density within the BWR core because the neutron moderator, as coolant, is boiling as it flows through the core which leads to two-phase flow.

Figure 3.3 shows a BWR core cross-sectional view and structures. There are some distinctive features to BWR core structures. One is the set of a cruciform control rod and four surrounding assemblies; the sets are regularly arranged in the core. Another feature is that the control rod and fuel assemblies are mutually separated by channel boxes wrapping each fuel assembly, in which fuel rods and water rods are tied in an array of, for example, 8 x 8 or 9 x 9, by spacers. A third feature is that the coolant flow path in the core is divided by the channel boxes and each fuel assembly makes a coolant flow path. There is a water gap region, in which water does not boil, between adjacent fuel assem­blies. The coolant flowing in from the core bottom comes to boiling and then a two-phase flow of water and steam is produced due to heat from the fuel rods within the fuel assemblies isolated by the channel boxes, but coolant is maintained in the non-boiling state outside the assemblies.

Those features of BWR core structures make it possible to improve or modify the intra-assembly design such as fuel rod size and array without changing specifications concerning the whole core construction such as control rod arrangement and fuel assembly shape.

Table 3.2 History of BWR Specifications [1, 2]

Type

BWR-l

BWR-2

BWR-3

BWR-4

BWR-5

GE

Improved Japan

BWR-6

ABWR

Construction/ USA

1956/1960

1964/1969

1966/1970

1967/1974

1973/1982

1974/1985

operation Japan

start year

1966/1970

1966/1971

1969/1974

1973/1978

1979/1984

1991/1996

Plant example USA [Power,

Dresden-1 (180)

Oyster Creek (636)

Dresden-2

(850)

Browns Ferry-1 (1155)

Lasalle-1 (1138)

Grand Gulf-1 (1266)

MWeJ Japan

Tsuruga-1 (350)

Fukushima

1-1

(460)

Fukushima 1-2

(780)

Tokai-2 (1100)

Fukushima II-2

(110)

Kashiwazaki Kariwa-6 (1350)

Features

First commer­cial power plant

Forced circulation Type of single direct cycle

Flow rate control by no. of pump rotations

Jet pump

Improvement in power density

Recirculation flow rate control by valves or M-G set MARK-II PCV

Improvement in operation per­formance

Reduction in radia­tion exposure Improvement in available factor

Improvement in power MARK-III PCV

Internal pumps Advanced control ele­ment drive mecha­nism RCCV

Fuel assembly type

6×6

7×7

7×7

7×7

8×8

8×8

8×8

9×9

Average core power density [kW/1] Forced circulation type

31 34 External loops(3-5 loops)

41 51 External loops (2 loops)

51

51

~54

51

Internal recirculation

of core coolant

External recirculation pump

External recirculation pump + jet pump

External recirculation pump

+ 5-nozzle jet pump

pumps

Containment vessel

Dry spherical type

MARK-I (flask type)

(Cone type) MARK-II

MARK-II advanced type

(Cylinder

type)

MARK-II

RCCV

Table 3.3 BORAX and early test BWRs

Test reactor

Research and development

BORAX-I

Verification of inherent safety and power excursion

BORAX-II

Research on pressurization and instability

BORAX-III

Electricity generation test

BORAX-IV

Examination on UO2 fuel use, stability, water radiolysis, and turbine radioactivity

BORAX-V

Examination on high power density core and nuclear superheat

EBWR (ANL, 5 MWe)

Power test and trouble experience

VBWR (GE)

Economical improvement test, natural/forced circulation, direct/indirect cycle, material test, etc.

image356

Fig. 3.3 BWR core cross-sectional view and structures