JPH0823596B2 - Primary circulation loop Water level meter Pressure water reactor - Google Patents

Description

TECHNICAL FIELD The present invention relates to a pressurized water nuclear reactor with improved safety in the event of an accident. Specifically, the present invention relates to various small break coolant loss accidents in a pressurized water reactor (PWR) including an accident at the Three Mile Island Nuclear Power Plant Unit 2 (TMI-2) in the United States, and subsequent plant recovery operations. When the primary circulation pump is stopped, the amount of water retained in the upper part of the reactor vessel that cannot be detected by the pressurizer water level meter alone is stored in the corresponding height part of the primary circulation loop without installing a water level meter in the reactor vessel. The present invention relates to a pressurized water reactor with improved safety in the event of an accident, detected by a differential pressure type water level gauge installed.

(Prior Art) Since the TMI-2 reactor accident was a loss of coolant from the top of the pressurizer (pressure relief valve), which is the only water level gauge installed device in the primary system, the primary system pressure decreases while At this point, the water level of the pressurizer increased, and the embarrassed reactor operator placed importance on the water level of the pressurizer and decided that there was enough water in the reactor. Therefore, in reality, the decrease in the amount of water held in the reactor vessel was not noticed for a long time, and other causes were also added, which eventually led to the worst situation of core melting.

Following the accident, the US Nuclear Regulatory Commission paid attention to this fact and instructed an improvement to directly detect the amount of water held in the reactor vessel of a pressurized water reactor. For this reason, Combation Engineering Co. (CE) added a heating thermocouple type water level gauge to the upper part of the reactor vessel, and Westinghouse (W company) added a differential pressure type water level gauge to the top and bottom of the reactor vessel. We have developed a system for detecting the amount of water held (see Reference 1).

However, this involves a structural strength problem of installing a measurement nozzle in the reactor vessel itself, and, as will be shown later, compared to the water level at the water level gauge installation position in the present invention, the mixed water level swelling due to core heat generation Therefore, there is a problem that the time when the water level drops below the position of the high temperature side pipe nozzle is considerably delayed, and there are certain restrictions on the installation, maintenance, adjustment, etc. of the water level gauge because it is located close to the core under high radiation. There is a problem.

(Problems to be Solved by the Invention) An object of the present invention is to solve the above problems and to provide a primary circulation loop water level gauge-added pressurized water reactor which facilitates detection of the amount of water held by the reactor.

(Means for Solving the Problem) In the present invention, the water level gauge is a WWR type PWR (W type PW).
R) or CE company type PWR and is installed in the same type PWR,
It is possible to eliminate or reduce the problems and restrictions of the water level gauge between the top and bottom of the reactor vessel. In the present invention, the water level gauge is an existing differential pressure type water level gauge with temperature compensation, and the representative density of the liquid is determined by the liquid temperature of the loop seal part and the representative pressure of the primary system, and the steam generator outlet plenum of the primary loop is obtained. Install between the upper end and the lower end of the loop seal. By installing in all loops, the surplus property is secured, and even if there are fluid or pressure oscillations in the primary system, the reliability of detection of the amount of water retained is performed by averaging the water level data over time and averaging the entire system. Can be raised.

Specifically, below the pressurizer water level gauge, in the portion from the steam generator outlet plenum including the vertical portion of the primary circulation loop above the core lower end to the loop seal lower end, a differential pressure type water gauge with temperature compensation is installed, It is used to detect the primary circulation flow rate during normal operation, and if there are various small fracture coolant loss accidents, including the United States Three Mile Island reactor accident, and the water volume in the reactor vessel drops when the primary circulation pump is stopped, the water level is reduced. Based on the characteristics that the meter can detect the water level drop preceding the core core water level drop and also detect the water level behavior corresponding to the water level at the upper part of the reactor vessel, The amount of water held in the upper part of the reactor vessel is detected, and similar water level gauges are installed in all loops of the primary circulation system to secure its surplus, and it is easy to secure the amount of water held necessary for removal of reactor decay heat. is there

 The pressurized water reactor of the present invention will be described with reference to the drawings.

FIG. 1 shows a W-type PWR, a pressurizer water level measurement range, and a water level gauge measurement range in the present invention. The main measurement range of the water level gauge in the present invention is about 2.2 m from the upper end of the steam generator outlet plenum to the lower end of the primary horizontal pipe section,
By including 3.7m below the lower end of the loop seal part in the measurement range, it is possible to detect the temporary reactor core water level lowering phenomenon at the time of loop seal clearing described later.

Fig. 2 shows a large-scale unsteady experimental device (LS) that simulates a W-type PWR (heat output 3423Mwt, 4 loops) at a volume ratio of 1/48 and actual height.
TF) shows the relationship between the height and the internal volume, and the measurement range of the water level gauge and the range of the core in the present invention. The relationship between the ratio of the volume of the primary system to the total volume and the height is almost the same as the W-type PWR, except for the pressurizer shortened by this device and the upper part of the steam generator outlet plenum where internal packing is removed. To do. This relationship expresses the relationship between the water level and the water volume in a calm state where no water level difference occurs in each part in the primary system, and is one guideline in the event of an accident. The pressurizer water level is higher than the reactor vessel, and the primary system volume corresponding to the measurement range is approximately 60% or more.

On the other hand, the primary system volume corresponding to a range of about 2.2 m above the lower end of the primary system horizontal pipe in the present invention is
It is in the range of 27-60%. Since 1/3 of the total volume is concentrated in this measurement range, the water level change in this range has the slowest time change compared to other parts in the primary system. The primary system volume corresponding to about 3.7 m below the primary system horizontal piping part is 15% of the total, and the total primary system volume of about 1/2 corresponds to the entire measurement range of the water level gauge in the present invention. are doing. The core portion is at a position corresponding to a volume of 7 to 20% of the total volume of the primary system, with the lower end of the loop seal portion being substantially the center thereof. It should be noted that the differential pressure introducing pipe of the water level gauge in the present invention is a thin tube of about 1/2 inch pipe, and even if one of the pipes is supposed to be broken, the flow rate flowing out from the pipe is high. (Or high-pressure filling system) can be sufficiently covered.

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

The LSTF (see Reference 2) can measure a differential pressure of about 4.5 m between the lower part of the steam generator outlet plenum and the lower end of the loop seal part. Among the small-breakage coolant loss accident simulation experiments conducted with this device, the SB3 experiment simulating the TMI-2 reactor accident and the same fracture area as the fracture position, the fracture position was the bottom of the reactor vessel and the low temperature side piping. , High temperature side piping, four experiments set on the top of the reactor vessel (SPI, SC respectively)
From the results of C, SH3, and SP2 experiments) (refer to Material 3), as shown in FIG. 3, it corresponds to the water level gauge in the present invention prior to the start of the fuel rod surface temperature rise due to the water level drop in the core. It can be seen that a differential pressure gauge response has occurred. That is, the water level starts to drop in the lower part of the steam generator outlet plenum at 408 seconds in the SPI experiment, 2088 seconds in the SP2 experiment, 1385 seconds in the SH3 experiment, and earlier than the core temperature rise start.

Since 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 of the present invention, the use of the water level gauge of the present invention allows the water level drop to be detected earlier than these experimental results. It is clear that The elapsed time from the start of lowering the water level at the steam generator outlet plenum to the lowering of the core water level varies depending on the fracture position in the primary system, as shown in Fig. 3, but naturally also depends on the fracture area. In the case of the LSTF experiment, this elapsed time was 981 seconds in the above-described SCC experiment, but in the experiment in which the fracture area was 10 times at the same fracture position, the elapsed time was 300 seconds.

Also, from FIG. 3, it can be seen that the water level in the steam generator outlet plenum starts to drop 300 to 1000 seconds earlier than the time when the water level in the reactor vessel begins to drop below the primary horizontal pipe nozzle position. This is because the fluid at the steam generator outlet plenum and the loop seal part is saturated water cooled by the steam generator secondary system, but on the core side, the voids generated by the decay heat of the core rise and steam This is because the drop in the mixed water level is delayed due to the flow to the generator. In the case of differential pressure type water level measurement in the reactor vessel, in addition to the influence of such a core upstream flow and void fraction distribution, the time when the water level drops below the high temperature side piping nozzle position in the present invention. This is later than in the case of the water level gauge, and therefore the elapsed time until the core water level is lowered becomes shorter than in the case of the present invention.

Next, the characteristics of the water level behavior on the descending side of the loop seal will be shown. In the SH3 experiment described above, the water level lowered from the steam generator outlet plenum was held at the same height for a long time after reaching the low temperature side horizontal piping position on the loop seal descending side, but during that time, the core The water level has dropped.
In other words, the retained water in the reactor vessel decreased with water remaining in the loop seal.

On the other hand, in the case of SP1, SP2 and SH3 experiments, the water level lowered from the steam generator outlet plenum was lower than the horizontal part of the low temperature side pipe, but did not reach the lower end of the loop seal, Position and after high pressure injection system operation,
It fell to the bottom. The core water level dropped during this period. SCC
In the case of the (pipe breakage on the low temperature side) experiment, the water level behavior was slightly different from these. The water level, which dropped from the steam generator outlet plenum to the horizontal part of the low temperature side pipe, continued to drop and reached the lower end of the loop seal part. For this reason, the steam on the descending side of the loop seal flowed into the rising side, and the retained water remaining in the loop seal part partially flowed into the reactor vessel from the low temperature side pipe. During this process, the core was temporarily exposed, and the temperature rise of the fuel rods above the central part of the core was detected, but with the temporary elimination of the pressure difference before and after the loop seal part,
The water level difference between the core in the reactor vessel and the downcomer also decreased, and the core was cooled temporarily. After this loop seal / clearing phenomenon, the core was exposed again due to the reduction of water content. These results show that the water level in the loop seal part below the horizontal part of the primary system pipe does not generally correspond to the water level change on the core side, but the water level drop from the steam generator outlet plenum to the horizontal part of the primary system pipe does not occur. This is useful for predicting water level drop on the core side. However, when the water level on the loop seal descending side reaches the lower end, as in the case of pipe breakage on the low temperature side, it is expected that the fuel rod temperature will rise due to the decrease in the core water level.

Finally, from the characteristics of the differential pressure gauge of the loop seal part (downward side) of the LSTF, the effect of pressure loss due to flow resistance in the pipe when the primary circulation pump is rotating is shown. In this differential pressure measurement range, there are no structures that obstruct the flow, and the pressure loss during this is negligibly small compared to the pressure loss between the top and bottom of the reactor vessel whose core is the measurement range. LSTF
In this case, the resistance coefficient was 2.75 × 10 ^-3 for the entire measurement range. Under the rated flow condition of an actual reactor, the pressure loss is comparable to that of the head, but when a water single-phase flow with a flow velocity of 2 m / s or less flows, the effect of the pressure loss can be ignored. Compared to the loop seal pipe of the actual furnace, the LSTF pipe has the same length and the diameter of the pipe is about 1/5, so it is considered that the resistance coefficient of this pipe is considerably larger than that of the actual furnace. In any case, when the water level gauge according to the present invention is installed in an actual reactor and used in a normal operation state, the flow resistance coefficient of the water level gauge measurement range is obtained in advance, and the influence of the pressure loss due to the primary circulating water is measured. By separating, it can be used to confirm the flow rate. The water level meter can be used under the condition that the rotation of the primary circulation pump is stopped, and the function of the water level meter can be approximately retained when the flow velocity is 2 m / s or less.

Reference Material 1. Yih-Yun Hsu et al., Some Possible Wa
ys to Improve Nuclear Power Plant Instrumentation,
Nucler Safety, 22 (6) 1981 Reference Material 2. The ROSA-IV Group, ROSA-IV Large Sc
ale Test Facility (LSTF) System Description, JAERI
−M 84−237 (Jan.1985) Reference Material 3. M.Suzuki, Break Location Effects on P
WR Small Break LOCA Phenomena−−Inadequate Core C
ooling in Lower Plenum Break Test at LSFT−−, JAER
I-M 88-271 (Jan. 1989).

[Brief description of drawings]

FIG. 1 is an explanatory view showing the relationship between the concept of the W-type PWR and the position of the pressurizer and the water level gauge in the present invention. 1 …… Reactor vessel 2 …… Core 3 …… Pressurizer 4 …… Steam generator 5 …… Primary circulation pump 6 …… Loop seal descending side 7 …… Loop seal ascending side 8 …… Primary system horizontal part ( Low-temperature side piping, high-temperature side piping) 9 ... Water level gauge in the present invention 10 ... Pressurizer water level gauge Fig. 2 shows the relationship between the height and volume ratio (11) of the LSTF device simulating a W-type PWR. 12 …… Steam generator outlet plenum upper end 13 …… Primary system horizontal pipe lower end 14 …… Loop seal lower end 15 …… Core upper end 16 …… Core lower end 17 …… Main water level gauge Main measurement range 18 …… All water level gauge Measurement range 19 ... Reactor core Figure 3 shows the relationship between the fracture position and the time of the main event in the LSTF small fracture experiment corresponding to the actual reactor 2 inch tube. 20 …… Reactor vessel bottom rupture (SP1) 21 …… Low temperature side pipe rupture (SCC) 22 …… High temperature side pipe rupture (SH3) 23 …… Reactor vessel top rupture 24 …… TM1-2 type rupture 25 …… Steam generator outlet plenum water level drop 26 …… Reactor vessel water level / drop from horizontal piping 27 …… Core exposure (water level drop)

Claims (1)

[Claims] 1. A temperature compensating differential pressure type water level gauge is installed in a portion from a steam generator outlet plenum including a vertical portion of a primary circulation loop below a pressurizer water level gauge and above a core lower end to a loop seal lower end. However, the primary circulation loop water level gauge-added pressurized water reactor is characterized in that the water level meter is used when the circulation flow is stagnant after the primary circulation pump is stopped.