Induced nuclear reactions: processes – systems – and elements – Reactor protection or damage prevention – By minimizing positive coolant void coefficient
Reexamination Certificate
1999-08-19
2002-01-29
Jordan, Charles T. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Reactor protection or damage prevention
By minimizing positive coolant void coefficient
C376S277000, C376S304000, C376S370000
Reexamination Certificate
active
06343106
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
This invention relates to nuclear reactors which circulate cooling water by natural circulation power, and to nuclear reactors which circulate at most 30% of the cooling water by forced-circulation power.
This invention improves critical power performance at the time of a pressure rise transient (rather than critical power performance at the time of a water-supply temperature transient) by making a dynamic void coefficient greater than a predetermined value.
2. Description of the Related Art
A conventional fuel assembly and a conventional reactor core are explained with reference to FIG.
13
and FIG.
14
.
FIG. 13
is a sectional view showing a principal part inside a conventional reactor core, and shows the arrangement of a fuel assembly and a control rod. In a fuel assembly
1
, two water-rods
5
where coolant (i.e. non-boiling water) flow inside are arranged and, for example, two or more fuel rods
3
containing nuclear fuel material
6
(like a cylinder pellet) are enclosed as shown in this figure. A channel box
4
surrounds the fuel assembly
1
, and a boiling water stream path
8
is formed in the channel box
4
.
The fuel assembly
1
and a control rod
2
are arranged regularly, in the reactor core
14
. A control rod
2
, for example, is arranged four per fuel assembly
1
.
The gap between a fuel assembly and an adjacent assembly serves as a by-pass portion
7
in which the coolant (i.e. non-boiling water) exists. The cooling water in the by-pass portion
7
does not boil, even if there is heating by neutron radiation, and the like.
The parameters d
1
and d
2
in
FIG. 13
represent the width of a channel box
4
, and a width of the by-pass portion
7
, respectively. The width d
2
of the conventional by-pass portion
7
is less than 10% of the width d
1
of the channel box
4
. In the conventional fuel assembly, the water rod(s)
5
is arranged at the central section of the fuel assembly
1
.
The following method is adopted in regard to the enrichment distribution of the fuel. The enrichment distribution is arranged in the vertical direction. The enrichment of the upper part is set 0.2 wt % (or less) larger than that of the lower part. There may be installed a natural uranium blanket of a low reactivity at a vertical edges of the fuel. In this case, the upper and lower blankets are the same length, or the upper blanket is longer than the lower blanket.
FIG. 14
is a system figure showing the outline of the conventional boiling water reactor. As shown in this figure, the nuclear reactor core
14
is built in the pressure vessel
11
of the boiling water reactor. The upper part of the nuclear reactor core
14
is surrounded with a shroud
13
. A steam separator
12
to separate steam and saturated water is installed in the upper part of the shroud
13
. By the steam separator
12
, the steam is drawn up and the saturated water is led to the outside of the shroud
13
.
Underneath the nuclear reactor core
14
, a plurality of control rods
15
are installed. The control rods
15
are driven in and out with control rod drive
16
installed in the lower part of the pressure vessel
11
.
Usually, the internal pressure of the pressure vessel
11
during operating power is set as 70 times atmospheric pressure, for example. The shroud
13
divides coolant flow to the upper part of the pressure vessel from the nuclear reactor core
14
(the arrow with attached symbol
10
a
in the figure) and flow exterior of the nuclear reactor core
14
, i.e., the coolant flow to the lower portion of the pressure vessel produced near the inner wall of the pressure vessel
11
(the arrow with attached symbol
10
b
in the figure).
Based on differences in the mechanisms which cause the flow of such coolant, there are two kinds of boiling water reactors, that is, a natural-circulation reactor and a forced-circulation reactor. In the natural-circulation reactor, cooling water is driven by natural-circulation power by the saturated water outside of the shroud
13
. Thus, the saturated water is led inside to the lower part of the nuclear reactor core
14
.
In contrast, the saturated water circulates in a forced-circulation water reactor by a drive from a power apparatus installed outside of the shroud
13
. As the power apparatus, there are employed a re-circulating water pump, an internal pump, etc.
A main steam pipe
17
is connected to the pressure vessel
11
, and the steam generated by the nuclear reactor is led to a high-pressure turbine
20
. A plurality of relief safety valves
18
are set in the main steam pipe
17
. When an abnormal pressure rise happens, a relief safety valve
18
is opened and the internal pressure of the pressure vessel
11
is reduced.
A turbine governor valve
19
is set between the high-pressure turbine
20
and the relief safety valves
18
to adjust the amount of steam introduced to the high-pressure turbine
20
.
When a so-called pressure rise transient phenomena occurs, such as during a loss of a generator load, in order to prevent the rotational frequency of the turbine from going too high, the turbine governor valve
19
is closed. When the turbine governor valve
19
is closed, the main steam is usually led to a condenser
23
through a by-pass line
28
.
A low-pressure turbine
21
is installed downstream of the high-pressure turbine
20
, and rotation of the turbines is converted into current by a generator
22
installed downstream of the low-pressure turbine
21
. The steam working in the turbine passes through a steam extraction line
29
a
from the low-pressure turbine
21
, is led to the condenser
23
, and is liquefied.
Through a feed pipe
27
and a feed pump
26
, the cooling water (i.e. condensation) made by the liquefying is returned to the pressure vessel
11
, and circulates through it. In the feed pipe
27
, a low-pressure feed water heater
24
and a high-pressure feed water heater
25
are provided. These feed water heater
24
and
25
heat the condensation to suitable water-supply temperature conditions. Feed water heater
24
and
25
operate by taking the steam extraction from each stage as a heat source to heat the coolant water (i.e. condensation) to the appropriate temperature condition. That is, heating of the condensation is performed by the steam extraction from the low-pressure turbine
21
and the high-pressure turbine
20
through steam extraction lines
29
b
and
29
c
, respectively, in the low-pressure feed water heater
24
and the high-pressure feed water heater
25
. The cooling water temperature of the outlet of the high-pressure feed water heater
25
is about 70 subcool temperature.
The boiling water reactor ensures a stability margin while operating under the following circumstances, by considering beforehand the worst conditions, such as a change in (&Dgr;MCPR) of the minimum critical power ratio (MCPR):
1) the internal pressure of the pressure vessel
11
experiences an abnormal rise;
2) an unusual transient change during operating power occurs in the form of a water supply temperature change outside of the normal range.
MCPR represents the minimum value of the ratio between fuel assembly power (critical power) expected when the boiling transition begins to happen and actual output. When the boiling transition begins to occur, since a liquid layer covering the fuel rod surface will be lost, the cooling state of the fuel rod surface will get worse, and the fuel temperature will rise.
Pressure rise transients occur due to, for example, loss of a load, and water-supply temperature transients occur due to improper operation of a temperature control unit and the like.
In the forced-circulation water reactor, although a flow rate transient due to failure of forced-circulation equipment is assumed, such a failure does not exist in a natural-circulation water reactor.
In connection with the pressure rise transient, the design assumes introduction of the steam to the by-pass line is successful. Generally, since the phenomena at failure is worst case, a stability margi
Foley & Lardner
Jordan Charles T.
Kabushiki Kaisha Toshiba
Richardson John
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