JPH02147990A - Reactor core flow measurement method - Google Patents
Reactor core flow measurement methodInfo
- Publication number
- JPH02147990A JPH02147990A JP63300517A JP30051788A JPH02147990A JP H02147990 A JPH02147990 A JP H02147990A JP 63300517 A JP63300517 A JP 63300517A JP 30051788 A JP30051788 A JP 30051788A JP H02147990 A JPH02147990 A JP H02147990A
- Authority
- JP
- Japan
- Prior art keywords
- differential pressure
- flow
- flow rate
- pump
- core
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
- 238000000691 measurement method Methods 0.000 title 1
- 238000005192 partition Methods 0.000 claims abstract description 7
- 238000000034 method Methods 0.000 claims description 7
- 239000002826 coolant Substances 0.000 claims 1
- 238000005259 measurement Methods 0.000 abstract description 17
- 238000012360 testing method Methods 0.000 abstract description 11
- 238000010586 diagram Methods 0.000 description 9
- 230000000694 effects Effects 0.000 description 3
- 239000000446 fuel Substances 0.000 description 3
- 239000000498 cooling water Substances 0.000 description 2
- 230000001105 regulatory effect Effects 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
【発明の詳細な説明】
〔産業上の利用分野〕
本発明は、原子炉圧力容器内部に冷却水の循環ポンプを
内蔵した原子炉に係り、特に、炉心流量測定精度の向上
に好適な原子炉内炉心流量測定方法に関する。[Detailed Description of the Invention] [Industrial Application Field] The present invention relates to a nuclear reactor having a cooling water circulation pump built into the reactor pressure vessel, and particularly to a nuclear reactor suitable for improving the precision of core flow rate measurement. Relating to a method for measuring inner core flow rate.
原子炉圧力容器に、直接、取り付けられた複数の循環ポ
ンプをもつ原子炉内の冷却水循環流量測定装置は、特開
昭５８−１０６９２号公報に記載され、第７図に示すよ
うに、複数個の循環ポンプ４の出入口近傍にポンプ差圧
計８の導圧管の開口部１６を設立し、この測定値と駆動
用モータ６に取り付けられた速度計７からの測定値を利
用して求める方法である。この時、循環ポンプ４の流社
−差圧の相関関係は、予め試験装置により採取しておく
。A cooling water circulation flow measuring device in a nuclear reactor having a plurality of circulation pumps directly attached to the reactor pressure vessel is described in Japanese Patent Application Laid-Open No. 10692/1982, and as shown in FIG. In this method, an opening 16 of the impulse pipe of the pump differential pressure gauge 8 is established near the entrance and exit of the circulation pump 4, and this measurement value and the measurement value from the speedometer 7 attached to the drive motor 6 are used to obtain the pressure. . At this time, the correlation between flow rate and differential pressure of the circulation pump 4 is collected in advance using a testing device.
上記従来技術は試験装置で循環ポンプ４の差圧−流量特
性を求めていたが、試験装置は完全に実機を模擬するこ
とが出来ず、試験装置における差圧−流量特性と実機に
おける差圧−流量特性の差異が生じ炉心流量計測精度を
低下する原因となっていた。The above-mentioned conventional technology uses a testing device to determine the differential pressure-flow rate characteristics of the circulation pump 4, but the testing device cannot completely simulate the actual machine. Differences in flow characteristics occurred, causing a decrease in core flow measurement accuracy.
本発明の目的は、この試験装置と実機における差圧−流
量の差異を無くすることにより、炉心流量の計測精度の
向上を図ることにある。An object of the present invention is to improve the measurement accuracy of the core flow rate by eliminating the difference between the differential pressure and the flow rate between this test device and the actual machine.
上記目的は、実機において循環ポンプ４の差圧流量特性
を求めることにより達成される。The above object is achieved by determining the differential pressure flow characteristics of the circulation pump 4 in an actual machine.
実機において、差圧−流量特性を求めることにより、試
験装置と実機との間に生じていた差圧−流量特性誤差を
考慮する必要が無くなる。従って、炉心流量計測精度が
向上し、プラントの信頼性、安全性が向上する。By determining the differential pressure-flow rate characteristic in the actual machine, there is no need to consider the differential pressure-flow rate characteristic error that occurs between the test device and the actual machine. Therefore, the core flow rate measurement accuracy is improved, and the reliability and safety of the plant are improved.
以下、本発明の実施例を第１図ないし第６図により説明
する。Embodiments of the present invention will be described below with reference to FIGS. 1 to 6.
第１図は、炉心１７の上部に、流量測定エレメント１９
．及び、流量調整弁２０をもつ模擬シュラウドへラドキ
ャンプ１５を仕切板２の上部に設置した例である。FIG. 1 shows a flow measuring element 19 in the upper part of the core 17.
．． This is also an example in which the Radcamp 15 is installed on the upper part of the partition plate 2 in a simulated shroud having a flow rate regulating valve 20.
起動試験時に、ポンプ４を運転し回転計７によりポンプ
回転数を設定し、ポンプ部差圧をポンプ部差圧計８によ
り測定する。炉心流量は、仕切板２の上部に設置された
模擬シュラウドへラドキャップ１５の流量測定エレメン
ト１９から差圧導圧管１１ａ、ｂを用いて求めた差圧を
差圧計１２、差圧流量計１３を介して求めることが出来
る。この方法により、回転数−ポンプ部差圧−流量の関
係が実機において求まり、従来の炉心流量計ａ１１方法
で含まれていた。工場試験と実機の流量−揚程特性の誤
差が無くなり、炉心流量計測精度の向上が図られる。During the startup test, the pump 4 is operated, the pump rotation speed is set using the tachometer 7, and the pump section differential pressure is measured using the pump section differential pressure gauge 8. The core flow rate is determined by applying the differential pressure obtained from the flow measurement element 19 of the rad cap 15 to the simulated shroud installed on the upper part of the partition plate 2 using the differential pressure impulse pipes 11a and 11b using the differential pressure gauge 12 and the differential pressure flow meter 13. It can be found through By this method, the relationship between the rotation speed, the differential pressure at the pump section, and the flow rate can be determined in the actual machine, which was included in the conventional core flow meter A11 method. This eliminates the error between the flow rate and head characteristics between the factory test and the actual machine, and improves the accuracy of core flow rate measurement.
第２図は、炉心】７の内部に圧損を変化させることが出
来る模擬燃料集合体１４を入れておく、この模擬燃料集
合体に流量測定エレメント２］−を入れておき、この流
量測定エレメントより求めた流量により、流量−ポンプ
部差圧−回転数の関係を求める。Fig. 2 shows a simulated fuel assembly 14 that can change the pressure drop inside the reactor core [7], a flow rate measuring element 2]- is placed in this simulated fuel assembly, and Based on the obtained flow rate, the relationship between flow rate - differential pressure at the pump section - rotational speed is determined.
第３図は、セパレータスタンドパイプ部で各スタンドパ
イプを流れる流量を電磁流量計等のシステム抵抗に影響
を与えない流量測定装置を設けておき、炉心流量と、ポ
ンプ部差圧（シュラウドサポートレグ間）−回転数の関
係を求めることが出来る。Figure 3 shows how the flow rate flowing through each standpipe is measured at the separator standpipe by installing a flow rate measuring device such as an electromagnetic flowmeter that does not affect system resistance. ) - rotation speed relationship can be found.
第４図なしい第６図は、各々ピトー管を炉心下部、仕切
板外、炉心上部に設けておき、このピトー管を用いて炉
心流量を求め、炉心流量−ポンプ部差圧（シュラウドサ
ポートレグ間）−回転数の関係を求める方法である。In Figure 6 without Figure 4, pitot tubes are installed at the bottom of the core, outside the partition plate, and at the top of the core, and the core flow rate is determined using these pitot tubes. This is a method to find the relationship between rotation speed) and rotation speed.
各々の方法において、流量−ポンプ部差圧−回転数の関
係をあらゆるポンプ運転台数で求めることが出来るため
、ポンプトリップ条件における炉心流量計測精度も向上
することができる。In each method, the relationship between the flow rate, the differential pressure at the pump section, and the rotational speed can be determined for any number of pumps in operation, so that the accuracy of core flow rate measurement under pump trip conditions can also be improved.
なお、図中１は原子炉圧力容器、３は気水分離器、５は
シャフト、９はポンプ部差圧計演算器。In the figure, 1 is a reactor pressure vessel, 3 is a steam separator, 5 is a shaft, and 9 is a pump section differential pressure gauge calculator.
】、０はポンプ部差圧流量計、１８は炉心支持板、２３
は流量測定変換器、２４は炉心流量計、２５はスタンド
パイプ、２６はピトー管である。], 0 is the pump section differential pressure flow meter, 18 is the core support plate, 23
2 is a flow rate measurement converter, 24 is a core flow meter, 25 is a stand pipe, and 26 is a pitot tube.
本発明によれば、試験装置と実機の間に生じる差圧−流
量の間に生じる誤差を考慮する必要が無くなり、炉心流
量の計測精度向上の効果がある。According to the present invention, it is no longer necessary to take into account the error that occurs between the differential pressure and the flow rate that occurs between the test equipment and the actual machine, and there is an effect of improving the measurement accuracy of the core flow rate.
第１図は本発明の一実施例の系統図、第２図は本発明の
第二の実施例の系ａ ＩＭ、第３図は本発明の第三の実
施例の系統図、第４図は本発明の第四の実施例の系統図
、第５図は本発明の第五の実施例の系統図、第６図は本
発明の第六の実施例の系統図、第７図は従来例の系統図
である。
１・・・原子炉圧力容器、２・・・仕切板、３・・・気
水分離器、４・・・循環ポンプ、５・・シャフト、６・
・・駆動用モータ、７・・速度計、８・・ポンプ部差圧
計、９ポンプ部差圧計演算器、１０・・・ポンプ部差圧
流量計、１．１ａ、ｂ・・・流量測定エレメント差圧導
圧管、１２・・・流ｆｉ謂定エレメント部差圧計、１３
・・・差圧流量計、１４・・・模擬燃料集合体、１５・
・・模擬シュラウドキャップ、１６・・・ポンプ部差圧
ΔＩＩＩ定用導圧管、１７・・・炉心、１８・・・炉心
支持板、１９・・・流量測定エレメント、２０・・・流
量調整弁、２１・・・電磁流量計、２２・・・ピトー管
、２３・・・流量測定変換器。
２４・・炉心流量計、２５・・・スタンドパイプ、２６
第
図
第３図
第２因
第
■
第
図
第６図Fig. 1 is a system diagram of an embodiment of the present invention, Fig. 2 is a system diagram of a second embodiment of the invention, Fig. 3 is a system diagram of a third embodiment of the invention, and Fig. 4 is a system diagram of an embodiment of the invention. is a system diagram of a fourth embodiment of the present invention, FIG. 5 is a system diagram of a fifth embodiment of the present invention, FIG. 6 is a system diagram of a sixth embodiment of the present invention, and FIG. 7 is a system diagram of a conventional system. FIG. 2 is an example system diagram. DESCRIPTION OF SYMBOLS 1... Reactor pressure vessel, 2... Partition plate, 3... Steam-water separator, 4... Circulation pump, 5... Shaft, 6...
... Drive motor, 7. Speed meter, 8. Pump section differential pressure gauge, 9 Pump section differential pressure meter calculator, 10... Pump section differential pressure flow meter, 1.1a, b... Flow rate measurement element Differential pressure impulse pipe, 12... Flow fi element part differential pressure gauge, 13
... Differential pressure flow meter, 14 ... Simulated fuel assembly, 15.
... Simulated shroud cap, 16 ... Pump section differential pressure ΔIII constant pressure impulse pipe, 17 ... Core, 18 ... Core support plate, 19 ... Flow rate measurement element, 20 ... Flow rate adjustment valve, 21... Electromagnetic flow meter, 22... Pitot tube, 23... Flow rate measurement converter. 24...Core flow meter, 25...Stand pipe, 26
Figure 3 Figure 2 Cause ■ Figure 6
Claims (1)
を囲む仕切板と、前記原子炉圧力容器の下部に設けた循
環ポンプと、この冷却材出入口部に設けられたポンプ部
差圧計と、前記循環ポンプの回転数を検知する速度計と
、前記循環ポンプ部の差圧と、前記速度計とから流量を
演算するポンプ部差圧流量計とよりなる炉心流量測定装
置において、 前記循環ポンプ部差圧と流量の関係を実機で求めること
を特徴とする炉心流量測定方法。[Scope of Claims] 1. A reactor pressure vessel, a partition plate surrounding the reactor core in the reactor pressure vessel, a circulation pump provided at the lower part of the reactor pressure vessel, and a circulation pump provided at the coolant inlet/outlet portion of the reactor pressure vessel. The core flow rate is composed of a pump section differential pressure gauge, which detects the rotation speed of the circulation pump, and a pump section differential pressure flowmeter that calculates the flow rate from the differential pressure of the circulation pump section and the speed meter. A method for measuring a reactor core flow rate, characterized in that, in the measuring device, the relationship between the differential pressure in the circulation pump section and the flow rate is determined using an actual machine.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63300517A JPH02147990A (en) | 1988-11-30 | 1988-11-30 | Reactor core flow measurement method |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63300517A JPH02147990A (en) | 1988-11-30 | 1988-11-30 | Reactor core flow measurement method |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH02147990A true JPH02147990A (en) | 1990-06-06 |
Family
ID=17885772
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP63300517A Pending JPH02147990A (en) | 1988-11-30 | 1988-11-30 | Reactor core flow measurement method |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH02147990A (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2019025315A1 (en) * | 2017-08-03 | 2019-02-07 | Framatome Gmbh | Device and method for flow measurement in the reactor core of a boiling-water reactore |
| CN110444301A (en) * | 2019-08-13 | 2019-11-12 | 中国核动力研究设计院 | Simulate supercritical pressure transient condition experimental provision and experimental method |
-
1988
- 1988-11-30 JP JP63300517A patent/JPH02147990A/en active Pending
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2019025315A1 (en) * | 2017-08-03 | 2019-02-07 | Framatome Gmbh | Device and method for flow measurement in the reactor core of a boiling-water reactore |
| CN110444301A (en) * | 2019-08-13 | 2019-11-12 | 中国核动力研究设计院 | Simulate supercritical pressure transient condition experimental provision and experimental method |
| CN110444301B (en) * | 2019-08-13 | 2022-07-01 | 中国核动力研究设计院 | Experimental device and experimental method for simulating supercritical pressure transient working condition |
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