JPH02183198A - Pressurized water nuclear reactor with water level gauge for primary circulation loop - Google Patents

Abstract

PURPOSE:To obtain an improved reliability of a volume detection of a held water by providing water level gauges between upper ends of outlet plenums and lower ends of loop seals of steam generators in all primary loops. CONSTITUTION:A temperature compensated pressure difference typed water level gauge 9 is provided at a part which is located between an outlet plenum of a steam generator 4 containing a vertical portion of a primary circulating loop which is located underneath a pressurizer water level gauge 10 and above a lower end of a reactor core 2, and a lower end of a loop seal, and moreover the same typed water level gauges are provided to all loops of the primary circulation systems. A volume of the primary system corresponding to a 2.2m range above a lower end of a primary system horizontal piping of the water level gauge 9 stays within a range of 27 to 60% of the whole volume. As one third of the whole volume is concentrated to this measuring range, a level change in this range is the slowest change with an elapsing time, as compared with other parts within the primary system. Also, a primary system volume corresponding to a 3.7m portion underneath the primary system horizontal piping is 15% of the whole volume and around 1/2 of the whole primary system volume, in total, corresponding to a whole measuring range of the water level gauge 9.

Description

DETAILED DESCRIPTION OF THE INVENTION (Field of Industrial Application) The present invention relates to a pressurized wooden nuclear reactor with improved safety in the event of an accident. Specifically, the present invention is applicable to various small breakage loss of coolant accidents in pressurized wood reactors (PWRs), including the Three Mile Shimabara Prepower Plant 2 (TMI-2) accident in the United States, and subsequent plant recovery. During operation and when the secondary circulation pump is stopped, the amount of water held in the reactor vessel that cannot be detected only by the pressurizer water level gauge can be detected by measuring the corresponding high water level in the secondary circulation loop without installing a water level gauge in the reactor vessel. This relates to a pressurized wooden nuclear reactor that has improved safety in the event of an accident by detecting the water level using a differential pressure type water level gauge installed at the bottom of the reactor.

(Prior art) The TM I-2 nuclear reactor accident mentioned above was an accident in which coolant was lost from the top of the pressurizer (pressure relief valve), which is the only equipment installed with a water level gauge in the -second system, so the -second system pressure was While the water level in the pressurizer was decreasing, the water level in the reactor rose, and the reactor operators were perplexed and decided that the water level in the reactor was important and that there was enough water in the reactor. Therefore, the fact that the amount of water held in the reactor vessel was actually decreasing was not noticed for a long time, and other causes were also added, which ultimately led to the worst case scenario of core meltdown. .

The U.S. Nuclear Regulatory Commission took note of this after the accident,
Ordered improvements to enable direct detection of the amount of water held in the reactor vessel of pressurized wooden reactors. And Westinghouse (
Company W) added differential pressure type water level gauges to the top and bottom of the reactor vessel (see Document 1). However, this
There is a structural strength problem of installing light nozzles in the reactor vessel itself, and as will be shown later, the high-temperature side piping nozzle is lower than the water level at the water level gauge installation position in the present invention due to mixed water level swelling due to core heat generation. There are problems such as a considerable delay in the time when the water level drops below the current level, and furthermore, there are certain restrictions on the installation, maintenance, and adjustment of water level gauges because they are located close to the reactor core, which is under high radiation.

(Problems to be Solved by the Invention) An object of the present invention is to provide a pressure wooden nuclear reactor with a primary circulation loop water level gauge that solves the above-mentioned problems.

(Means for solving the problem) In the present invention, the water level gauge is a PWR (W type P
WR), and can eliminate or reduce the problems and limitations of water level gauges between the top and bottom of the reactor vessel. In the present invention, the water level gauge is an existing temperature-compensated differential pressure type water level gauge, and the representative density of the liquid is determined from the liquid temperature at the loop seal portion and the following representative pressure.
- Installed between the upper end of the steam generator outlet plenum and the lower end of the loop seal for all subsequent loops. By installing it in all loops, redundancy is ensured, and even if there are fluid or pressure fluctuations in the primary system, by averaging the water level data over time and averaging the entire system, reliability in detecting the amount of water held is ensured. You can increase the degree.

Therefore, the pressurized water reactor of the present invention has a temperature-compensated section from the steam generator outlet plenum to the lower end of the loop seal, including the vertical section of the primary circulation loop below the pressurizer water level gauge and above the lower end of the core. This is a Westinghouse-type pressurized water nuclear reactor that is equipped with a differential pressure water level gauge and a similar water level gauge installed in all loops of the primary circulation system.

Specifically, a temperature-compensated differential pressure water level gauge is installed in the area from the steam generator outlet plenum to the lower end of the loop seal, including the vertical part of the primary circulation loop below the pressurizer water level gauge and above the lower end of the core. During normal operation, it is used to detect the primary circulation flow rate, and when the amount of water held in the reactor vessel decreases due to various small fracture loss of coolant accidents, including the Three Mile Island nuclear reactor accident in the United States, or when the primary circulation bomb is stopped, the water level Based on the characteristic that the water level gauge can detect the drop in water level that precedes the drop in the water level in the reactor core, and can also detect the water level behavior corresponding to the water level inside the reactor vessel, it is possible to detect the water level in the reactor vessel without installing a water level gauge in the reactor vessel itself. A primary circulation loop water level gauge that detects the amount of water held within the reactor, and also installs similar water level gauges in all loops of the primary circulation system to ensure surplus, making it easy to secure the amount of water needed to remove reactor decay heat. It is a pressurized water reactor.

The pressurized water nuclear reactor of the present invention will be explained with reference to the drawings. Figure 1 shows the W-type PWR, the water level measurement range of the pressurizer, and the measurement range of the water level meter in the present invention.The main measurement range of the water level meter in the present invention is from the upper end of the steam generator outlet plenum to the primary system horizontal measurement piping. The measurement range is 2.2m to the bottom end, but by including 3.7m below that to the bottom end of the loop seal, it is also possible to detect the temporary core water level drop phenomenon during loop seal clearing, which will be described later. Can be done. Figure 2 shows a W-type PWR (thermal output 3423
The relationship between the height and internal volume in a large unsteady experimental facility (LSTF), which simulates M, 4 loops) at a volume ratio of 1/48 and the actual height, and the measurement range of the water level gauge and the range of the reactor core in the present invention. show. -The relationship between the ratio of the internal volume of the next system to the total volume and the height is 1.If you exclude the pressurizer, which has been made thicker and shorter in this device, and the upper part of the steam generator outlet plenum from which internal fillings have been removed, it is almost W-shaped. This corresponds to PWR.

This relationship represents the relationship between water level and water volume in a calm state where there is no difference in water level in each part of the subsystem, and can serve as a guide in the event of an accident. The pressurizer water level gauge is located at a higher position than the reactor vessel, and the primary system volume corresponding to its measurement range is approximately 60% or less.

On the other hand, in the present invention, the primary system volume corresponding to the 2.2 m range above the lower end of the primary horizontal piping of the water level needle is in the range of 27 to 60% of the total volume. Since 1/3 of the total volume is concentrated in this measurement range, the water level changes in this range are the most gradual over time compared to other parts of the primary system. - The volume of the primary system corresponding to the 3.7m below the horizontal piping section of the secondary system is 15% of the total, and the total volume is approximately 17m of the total.
The primary system volume of 2 corresponds to the entire measurement range of the water level meter in the present invention. The core portion is located at a position corresponding to 7 to 20% of the total volume of the -order system, with the lower end of the loop seal being approximately at its center.

In addition, the differential pressure impulse pipe of the water level gauge in the present invention is a thin pipe of about 172 inches, and even if one of these pipes were to break, the flow rate flowing out from it would be limited to the high pressure injection system (or high-pressure filling system).

(Example) Next, the present invention will be specifically explained with reference to Examples.

For LSTF (see Document 2), approximately 4.5 m between the bottom of the steam generator outlet plenum and the bottom end of the loop seal.
It is possible to measure the differential pressure at a height of . Among the simulation experiments of small fracture loss of coolant accidents conducted using this equipment, the SB3 experiment, which simulates the TM I-2 nuclear reactor accident, and the SB3 experiment, which had a similar fracture area and the fracture location was at the bottom of the reactor vessel, on the low-temperature side. From the results of four experiments (abbreviated as SPl, SCC, SH3, and SP2 experiments, respectively) set on piping, high-temperature side piping, and the top of the reactor vessel (refer to Document 3), the water level in the reactor core was determined as shown in Figure 3. It can be seen that the response of the differential pressure gauge, which corresponds to the water level gauge in the present invention, occurred prior to the start of the fuel rod surface temperature rise due to the decrease. In other words, the start of the water level drop at the bottom of the steam generator outlet plenum is 4 in the case of the SPI experiment.
08 seconds, 2088 seconds in the case of the SP2 experiment, and 1385 seconds in the case of the SB3 experiment, which is earlier than the start of the core temperature rise.

The upper end of this LSTF loop seal differential pressure gauge is located 1.4 m lower than the upper end of the water level gauge in the present invention, so if the water level gauge in the present invention is used, a drop in water level will be detected even earlier than these experimental results. It is clear that As shown in FIG. 3, the elapsed time from the start of the water level drop in the steam generator outlet plenum to the core water level drop varies depending on the position of the -order system fracture, but it also naturally depends on the fracture area. In the case of the LSTF experiment, this elapsed time was 981 seconds in the SCC experiment described above, but in an experiment with the same fracture location and 10 times the fracture area, the elapsed time was about 300 seconds.

Furthermore, from FIG. 3, it can be seen that the water level in the steam generator outlet plenum begins to fall 300 to 1000 seconds earlier than the time when the water level in the reactor vessel starts to fall below the position of the primary horizontal piping nozzle. This is because the fluid in the steam generator outlet plenum and loop seal is saturated water cooled by the steam generator secondary system, while on the reactor core side, voids generated by the decay heat of the core rise. This is due to the delay in lowering the mixing water level as it flows to the steam generator.

In the case of differential pressure type water level measurement in the reactor vessel, in addition to being affected by the upward flow in the core and the void fraction distribution, the present invention also determines the time when the water level drops below the high temperature side piping nozzle position. Therefore, the elapsed time until the core water level drops is shorter than in the case of the present invention.

Next, the characteristics of the water level behavior 7 on the lowering side of the loop seal will be described. In the above-mentioned SB3 experiment, the water level that dropped from the steam generator outlet plenum reached the low-temperature side piping (horizontal part) on the descending side of the loop seal, and then remained at the same height for a long time. During that time, the core water level decreased. In other words, the amount of water retained in the reactor vessel decreased while water remained in the loop seal.

On the other hand, in the case of SPI, SF3, and SH3 experiments, the water level lowered from the steam generator outlet plenum is lower than the horizontal part of the low-temperature side piping, but does not reach the lower end of the loop seal, and is located between this horizontal part and the lower end. position, and after activation of the high-pressure injection system, it dropped to the lower end. The core water level decreased during this time.

In the case of the SCC (cold side pipe rupture) experiment, water level behavior was slightly different from these. The water level that had dropped from the steam generator outlet plenum to the horizontal section of the cold-side piping continued to drop further and reached the lower end of the loop seal. Therefore, the steam on the descending side of the loop seal flowed into the ascending side, and a portion of the water remaining in the loop seal flowed into the reactor vessel from the low-temperature side piping. During this process, the reactor core was temporarily exposed and a temperature rise in the fuel rods above the center of the core was detected, but as the pressure difference before and after the loop seal was temporarily resolved, The water level difference between the core and the downcomer also decreased, and the core was temporarily cooled. After this loop-seal clearing phenomenon, the core was exposed again due to water retention. What is clear from these results is that the water level at the loop seal below the horizontal section of the primary system piping does not correspond to changes in the water level on the core side, but the water level drop from the steam generator outlet plenum to the horizontal section of the primary system piping does. This will be useful in predicting the drop in water level on the core side. However, if the water level on the descending side of the loop seal reaches the lower end, as in the case of a rupture of the low-temperature side pipe, it is expected that the temperature of the fuel rods will increase due to the drop in the core water level.

Finally, from the characteristics of the differential pressure gauge at the loop seal part (downward side) of the LSTF, we will show the influence of pressure loss due to flow resistance in the piping when the secondary circulation pump is rotating. There are no structures that obstruct the flow in this differential pressure measurement range, and the pressure loss during this time is limited to the top of the reactor vessel, which has the core in the measurement range.
Compared to the pressure loss between the bottoms, it is negligible. L
Resistance coefficient for STF is 2.75 for the entire measurement range
x 10-3. Under the rated flow conditions of an actual reactor, the pressure loss is comparable to the water head, but when a turbine-phase flow with a flow rate of 2 m/s or less flows, the effect of the pressure loss can be ignored. Compared to loop seal piping in an actual furnace, LST
Since the F piping is the same length and the piping diameter is approximately 115,
The resistance coefficient of this piping is considered to be considerably larger than that of an actual furnace. In any case, when installing the water level meter according to the present invention in an actual reactor and using it in normal operating conditions, the flow resistance coefficient of the water level meter measurement range should be determined in advance, and the influence of pressure loss due to circulating water By separating it, it can be used to check the flow rate. It can be used as a water level gauge under the condition that the rotation of the secondary circulation bomb is stopped, and when the flow velocity is 2 m/sec or less, the function of the water level gauge can be maintained approximately.

Reference materials 1. Yih-Yun Hsu et a.
l, Some Possible Ways to
Improve Nuclear Power P
lant Instrumentation, Nuc
lear 5afety+ 22(6) (1981
).

Reference materials 2. The ROSA-IV Grou
p, ROSA-IVLarge 5cale Te5
t Facility (LSTF) System
mDescription, JAERI-M
84-237 (Jan, 1985).

Reference materials 3. M, 5uzuki, Break
LocationEffects on PWR
Small Break LOCA Pheno
menaInadequate Core Cooli
ng in Loever Plenum Brea
kTest at LSTF--, JAERI-M(t
o be published).

[Brief explanation of the drawing]

FIG. 1 is an explanatory diagram showing the relationship between the concept of a W-type PWH and the positions of a pressurizer and a water level gauge in the present invention. 1. Reactor vessel 2, core 3, pressurizer 4, steam generator 5, secondary circulation pump 6, loop seal descending side 7, loop seal ascending side 8, secondary system piping horizontal section (low temperature side piping, high temperature side Piping) 9,
Water level gauge 10 and pressurizer water level needle in the present invention FIG. 2 shows the relationship (11) between the height and volume ratio of an LSTF device simulating a W-type PWR. 12, Steam generator outlet plenum upper end 13, secondary system piping horizontal lower end 14, loop seal lower end 15, core upper end 16, core lower end 17, Water level meter measurement range 18, Water level meter total measurement range 19, Core No. 3 The figure shows the relationship between the fracture position and the time of major events in the LSTF small fracture experiment and the equivalent of a 2-inch tube in an actual reactor. 20. Reactor vessel bottom rupture (SPI) 21° 22° 23° 24° 25° 26° 27° Cold side pipe rupture (SCC) High temperature pipe rupture (SH3) Reactor vessel top rupture TMI-2 type rupture steam generation Lower reactor vessel outlet plenum water level Lower reactor vessel water level/horizontal piping section exposed core (lower water level)

Claims (1)

[Claims] A temperature-compensated differential pressure water level gauge is installed in the area from the steam generator outlet plenum to the lower end of the loop seal, which includes the vertical part of the primary circulation loop below the pressurizer water level gauge and above the bottom end of the core. A Westinghouse type primary circulation loop water level gauge pressure water reactor characterized by having similar water level gauges installed in all loops of the system.