Induced nuclear reactions: processes – systems – and elements – Testing – sensing – measuring – or detecting a fission reactor... – Flux monitoring
Reexamination Certificate
2001-06-08
2002-06-18
Behrend, Harvey E. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Testing, sensing, measuring, or detecting a fission reactor...
Flux monitoring
C376S217000
Reexamination Certificate
active
06408041
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
The present invention relates to an in-core fixed nuclear instrumentation system and a power distribution Monitoring system of a reactor such as a boiling water type reactor.
2. Description of the Prior Art
A reactor, for example, a boiling water type reactor (hereinafter, referred simply to as BWR) is provided with a power distribution monitoring system which monitors a reactor operating mode and a reactor power distribution (hereinafter, in this specification, the reactor power distribution is described as “in-core power distribution”, “core power distribution”, or the like), as generally shown in FIG.
23
and FIG.
24
.
In the BWR, as shown in
FIG. 23
, a reactor pressure vessel
2
is housed in a primary containment vessel
1
, and a reactor core
3
is accommodated in the reactor pressure vessel
2
. As shown in
FIG. 24
, the reactor core
3
is constructed in a manner that many fuel assemblies
4
and control rods
5
are mounted therein. In the react or core
3
, an in-core nuclear instrumentation assembly
6
is located on a position in the reactor core
3
, which is surrounded by fuel assemblies
4
.
As shown in
FIG. 24
, the in-core nuclear instrumentation assembly
6
is arranged in a corner water gap G formed by four fuel assemblies
4
, and neutron detector s
8
are discretely arranged on several positions along a core e axial direction in a nuclear instrumentation tube (pipe)
7
.
The neutron detector
8
i s a so-called fixed type, and in a boiling water type reactor (BWR), usually, four neutron detectors are discretely arranged on a fuel effective portion at equal intervals.
Further, in the nuclear instrumentation tube
7
, a TIP (Traversing In-Core Probe) conduit pipe
9
is arranged, and in the TIP conduit pipe (tube)
9
, one traversing neutron detector (TIP)
10
is located so as to be movable in an axial direction of the TIP conduit pipe
9
. Moreover, as shown in
FIG. 23
, there is located a movable type neutron flux measuring system for axially continuously measuring neutron flux by means of a matrix device
11
, a TIP driving device
12
, a TIP control and neutron flux signal processing system
13
or the like. A reference numeral
14
denotes a penetration section, a reference number
15
denotes a valve mechanism and a reference number
16
denotes a shielding container. These neutron detectors
8
and
10
, control systems such as signal processing systems
13
and
17
(will be described later) for neutron detectors
8
and
10
constitute a reactor nuclear instrumentation system
18
.
On the other hand, the in-core fixed neutron detectors
8
{LPRM (local power range monitor) detector} arranged in the reactor core are divided into some groups, and then, an average signal {APRM (average power range monitor) signal} for each group is generated so that an power level of a power range of the reactor core
3
is monitored on the basis of these APRM signals. More specifically, when an abnormally transient phenomenon or accident such that neutron flux rapidly rises up happens, the LPRM detectors
8
detect the transient phenomenon and the occurrence of accident so that, according to the APRM signals generated by the detected signals of the LPRM detectors
8
, a reactor safety protection system (not shown) rapidly makes a scram operation of a reactor scram system (not shown) such as a control rod driving mechanism or the like in order to prevent fuel assemblies or the reactor core from being break down. That is, the LPRM detector
8
is constituted as a part of the reactor safety protection system.
By the way, in individual in-core fixed neutron detectors
8
, a sensitivity change takes place by neutron irradiation or the like. In order to calibrate a sensitivity of each neutron detector
8
for each predetermined period during operation, the TIP (traversing neutron detector)
10
is operated so as to obtain a continuous power distribution in a core axial direction, and the sensitivity change of each neutron detector
8
is corrected by means of a gain adjusting function of a neutron detector (LPRM) signal processing system
17
. A detection signal S
2
detected by the neutron detector
8
is processed by means of the signal processing system
17
, and thereafter, is transmitted to a process computer
20
which will be described later.
In general, the BWR is provided with a process control computer
20
for monitoring an operating mode and power distribution of a nuclear (atomic) power plant. The process control computer
20
is provided with a nuclear instrumentation control system
21
for monitoring and controlling the reactor nuclear instrumentation system
18
, a power distribution simulating system
22
including a physics model having three-dimensional thermal-hydraulics simulation code, and an input-output system
23
. The reactor power distribution simulating system
22
is incorporated in one or plural process control computers
20
as a program. Further, the reactor power distribution simulating system
22
includes a power distribution simulating module
24
and a power distribution learning (adaptive) module
25
.
Neutron flux signal obtained by the TIP
10
of the reactor nuclear instrumentation system
18
is processed as a nuclear instrumentation signal corresponding to a core axial direction position by means of the TIP neutron flux signal processing system
13
of the reactor nuclear instrumentation system
18
. Then, the nuclear instrumentation signal is read via the nuclear instrumentation control system
21
of the process control computer
20
into-the power distribution simulating system
22
as a reference power distribution in a three-dimensional nuclear thermal-hydraulics simulation.
On the other hand, core state data S
3
(process quantity) including a control rod pattern, a core coolant flow rate, an internal pressure of the reactor pressure vessel, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on, which are used as various operating parameters indicative of a reactor operating mode (state) and obtained from a core state data measuring apparatus
26
as a reactor core state data measuring means, is read into a core state data processing system
27
, and then, is processed so that a reactor thermal output or the like is calculated. Then, the reactor core state data S
3
including the calculated reactor thermal output is transmitted to the reactor power distribution simulating system
22
via the nuclear instrumentation control system
21
of the process control computer
20
.
In fact, the reactor core state data measuring apparatus
26
is composed of a plurality of monitoring devices. In addition, the reactor core state data measuring apparatus
26
is a general name of an apparatus for collecting process data of various operating parameters of the reactor, and is shown as one measuring apparatus in
FIG. 23
for simplification. Moreover, the core state data processing system
27
may be used as one function of the process control computer
20
.
The detection signals S
2
and the core state data S
3
transmitted in the aforesaid manner are transmitted to the power distribution simulating system
22
of the process control computer
20
. In the power distribution simulating system
22
, a core power distribution is simulated on the basis of the transmitted core state data S
3
and the three-dimensional nuclear thermal-hydraulics simulation code of the power distribution simulating module
24
. Further, the power distribution simulating system
22
learns a reference power distribution of the core nuclear instrumentation data by a learning function (adapting function) of the power distribution learning (adapting) module
25
, and then, corrects the simulation result (core power distribution) while referring to the reference power distribution. As a result, in a power distribution predictive simulation after that, it is possible to accurately simulate a reactor power distribution.
In the conventional in-core
Behrend Harvey E.
Kabushiki Kaisha Toshiba
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