JPS5949484B2 - Steam generator water level control device - Google Patents

Steam generator water level control device

Info

Publication number
JPS5949484B2
JPS5949484B2 JP3883575A JP3883575A JPS5949484B2 JP S5949484 B2 JPS5949484 B2 JP S5949484B2 JP 3883575 A JP3883575 A JP 3883575A JP 3883575 A JP3883575 A JP 3883575A JP S5949484 B2 JPS5949484 B2 JP S5949484B2
Authority
JP
Japan
Prior art keywords
steam generator
water level
steam
flow rate
feed water
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.)
Expired
Application number
JP3883575A
Other languages
Japanese (ja)
Other versions
JPS51113001A (en
Inventor
耕治 桑原
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Mitsubishi Heavy Industries Ltd
Original Assignee
Mitsubishi Heavy Industries Ltd
Priority date (The priority date 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 date listed.)
Filing date
Publication date
Application filed by Mitsubishi Heavy Industries Ltd filed Critical Mitsubishi Heavy Industries Ltd
Priority to JP3883575A priority Critical patent/JPS5949484B2/en
Publication of JPS51113001A publication Critical patent/JPS51113001A/en
Publication of JPS5949484B2 publication Critical patent/JPS5949484B2/en
Expired legal-status Critical Current

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  • Control Of Non-Electrical Variables (AREA)
  • Control Of Steam Boilers And Waste-Gas Boilers (AREA)

Description

【発明の詳細な説明】 本発明は原子力発電プラントにおける蒸気発生器の水位
制御装置に関する。
DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a water level control device for a steam generator in a nuclear power plant.

蒸気発生器を効率的にかつ安全に運転するためには、前
記蒸気発生器の水位を常に望ましい範囲に保持する必要
がある。
In order to operate a steam generator efficiently and safely, it is necessary to maintain the water level in the steam generator within a desired range at all times.

このため従来は、蒸気発生器出口側の主蒸気流量、ドラ
ム水位および蒸気発生器への給水流量を検出し、これら
の検出値に基づいて給水弁を操作して蒸気発生器の水位
を制御していた。
For this reason, conventionally, the main steam flow rate, drum water level, and water supply flow rate to the steam generator at the steam generator outlet side were detected, and the water level in the steam generator was controlled by operating the water supply valve based on these detected values. was.

しかしながらこのような従来型装置では、タービンが極
低負荷の場合、主蒸気流量および給水流量が小さく、そ
の正確な検出が困難であるため自動制御を行なうことが
できないという不具合があった。
However, such a conventional device has a problem in that when the turbine is under extremely low load, the main steam flow rate and the feed water flow rate are small, and it is difficult to accurately detect them, so that automatic control cannot be performed.

この解決策として、ドラム水位の検出値のみより水位制
御を自動で行なう方法も考え出されたが、低負荷時には
タービン抽気を摂ることかできないため、給水温度が低
く、しかも低負荷時の主蒸気温度が逆に高いという原子
力発電所の特性から、給水流量を増加すると、蒸気発生
器内に、大量の低温給水が運びこまれることとなる。
As a solution to this problem, a method was devised to automatically control the water level based only on the detected value of the drum water level, but since turbine bleed air cannot be taken in at low loads, the feed water temperature is low, and the main steam at low loads is Due to the characteristic of nuclear power plants that the temperature is conversely high, increasing the feed water flow rate will result in a large amount of low temperature feed water being carried into the steam generator.

この種の蒸気発生器においては、加熱用高温流体すなわ
ち原子炉冷却材の流量は、一定に保持され、かつ、その
温度も実質的に一定に保持されており、直ちに交換熱量
は変わらないから蒸気発生器内の水温が一時的に低くな
る。
In this type of steam generator, the flow rate of the high-temperature heating fluid, that is, the reactor coolant, is held constant and its temperature is also held substantially constant, so the amount of heat exchanged does not change immediately, so the steam The water temperature in the generator becomes temporarily low.

蒸気発生器内は、一般に蒸気泡のない下部の加熱領域と
、蒸気泡が発生し気相と液相の二相状態にある上部の沸
騰領域に分けて観念されるが、前述のように蒸気発生器
内の水温が低くなると、前記2領域の境界が上方へ移動
し、かつ沸騰領域の容積が減少する。
The inside of a steam generator is generally divided into a lower heating area where there are no steam bubbles, and an upper boiling area where steam bubbles are generated and a two-phase state of gas and liquid phases. As the water temperature in the generator decreases, the boundary between the two zones moves upward and the volume of the boiling zone decreases.

つまりかえって水位が低下するといういわゆる逆応答が
生じる。
In other words, a so-called reverse response occurs in which the water level decreases.

このため負荷変動に追従しないという新たな不具合が発
生するので、このような方法は長時間一定負荷運転以外
には用いられなかった。
This causes a new problem of not being able to follow load fluctuations, so this method has not been used for anything other than long-term constant load operation.

本発明は、上記従来装置の不具合に鑑みなされたもので
、蒸気発生器の蒸気出口側に主蒸気温度検出器を、前記
蒸気発生器の給水入口側に給水温度検出器をそれぞれ設
け、第1の演算回路で前記主蒸気温度検出器の出力と前
記給水温度検出器の出力とから前記蒸気発生器の中での
伝達熱量を算出し、さらに原子炉出力検出器の出力と前
記伝達熱量とから第2の演算回路において、給水流量設
定値を算出し、同給水流量設定値により、前記蒸気発生
器の給水入口に連通ずる管に設けられた給水弁を操作し
て、前記蒸気発生器の水位制御を自動的に行なうもので
、本発明による主蒸気および給水の温度計測は容易かつ
正確であり、タービンの低負荷領域において蒸気発生器
の自動水位制御を可能にしたものである。
The present invention was made in view of the above-mentioned problems of the conventional device, and includes a main steam temperature detector provided on the steam outlet side of the steam generator and a feed water temperature detector provided on the feed water inlet side of the steam generator. An arithmetic circuit calculates the amount of heat transferred in the steam generator from the output of the main steam temperature detector and the output of the feed water temperature detector, and further calculates the amount of heat transferred in the steam generator from the output of the reactor power detector and the amount of transferred heat. A second calculation circuit calculates a water supply flow rate setting value, and uses the water supply flow rate setting value to operate a water supply valve provided in a pipe communicating with the water supply inlet of the steam generator to raise the water level of the steam generator. Control is performed automatically, and the temperature measurement of main steam and feed water according to the present invention is easy and accurate, and enables automatic water level control of the steam generator in the low load region of the turbine.

以下本発明の実施例を図面にもとづいて説明する。Embodiments of the present invention will be described below based on the drawings.

図面において原子炉冷却材の循環系統1は、循環ポンプ
2、原子炉3、蒸気発生器4および配管5.6.7によ
り構成されている。
In the drawing, a reactor coolant circulation system 1 is composed of a circulation pump 2, a nuclear reactor 3, a steam generator 4, and piping 5.6.7.

これに対し給水若しくは蒸気の循環系統8は、給水ポン
プ9、給水弁10.蒸気発生器4.タービン11および
配管12,13,14,15により構成されている。
On the other hand, the water supply or steam circulation system 8 includes a water supply pump 9, a water supply valve 10. Steam generator 4. It is composed of a turbine 11 and pipes 12, 13, 14, and 15.

前記配管13には給水弁10に近接して流量計16.蒸
気発生器4に近接して給水温度検出器17が夫々配設さ
れている。
A flow meter 16 is installed in the pipe 13 adjacent to the water supply valve 10. Feed water temperature detectors 17 are disposed close to the steam generators 4, respectively.

さらに配管14には蒸気発生器4に近接して主蒸気温度
検出器18が設けられている。
Furthermore, a main steam temperature detector 18 is provided in the piping 14 in close proximity to the steam generator 4 .

一方、原子炉3には原子炉出力器19、蒸気発生器4に
は水位検出器20が夫々配設あれている。
On the other hand, the reactor 3 is provided with a reactor power device 19, and the steam generator 4 is provided with a water level detector 20, respectively.

前記給水温度検出器17および前記主蒸気温度検出器1
8はそれぞれ独立して同一の第1の演算回路21に電気
的に結ばれていて、さらに同第1の演算回路21および
前記原子炉出力検出器19もそれぞれ独立して同一の第
2の演算回路22に電気的に結ばれている。
The feed water temperature detector 17 and the main steam temperature detector 1
8 are each independently electrically connected to the same first arithmetic circuit 21, and furthermore, the first arithmetic circuit 21 and the reactor power detector 19 are also independently connected to the same second arithmetic circuit 21. It is electrically connected to circuit 22.

同第2の演算回路22は、補償回路23に電気的に結ば
れ、さらに同補償回路23は比較回路24を介して前記
水位検出器20に電気的に結ばれると共に前記流量計1
6および制御器25に各別に電気的に結ばれている。
The second arithmetic circuit 22 is electrically connected to a compensation circuit 23, and the compensation circuit 23 is further electrically connected to the water level detector 20 via a comparison circuit 24, and the flowmeter 1
6 and the controller 25, respectively.

そして同制御器25は前記給水弁10に電気的に結ばれ
ている。
The controller 25 is electrically connected to the water supply valve 10.

上記した構成の本実施例において原子炉冷却材は循環系
統1を矢印の方向に流れ、このとき原子炉3内で熱を受
けとり蒸気発生器4内で循環系統8を流れる給水を加熱
し、蒸気を発生させる。
In this embodiment with the above configuration, the reactor coolant flows through the circulation system 1 in the direction of the arrow, and at this time, it receives heat in the reactor 3, heats the feed water flowing through the circulation system 8 in the steam generator 4, and generates steam. to occur.

これに対応し給水は配管13から前記蒸気発生器4内に
流入し前記原子炉冷却材から受熱し蒸気となって配管1
4に流出する。
Correspondingly, the feed water flows into the steam generator 4 from the piping 13, receives heat from the reactor coolant, becomes steam, and enters the piping 13.
4.

前記したように蒸気発生器4内で熱量の授受が行われる
が、その伝達熱量は、給水温度検出器17によって給水
温度を検出し同時に蒸気発生器4から流出する蒸気いわ
ゆる飽和蒸気である主蒸気の温度を主蒸気温度検出器1
8によって検出し、前記雨検出値を第1の演算器21に
送ってここで演算することにより算出される。
As described above, the amount of heat is exchanged within the steam generator 4, and the amount of transferred heat is determined by the temperature of the feed water being detected by the feed water temperature detector 17, and at the same time, the steam flowing out from the steam generator 4 is the main steam, which is saturated steam. Main steam temperature detector 1
8, and the rain detection value is sent to the first arithmetic unit 21, where it is calculated.

この第1の演算器21の出力信号を第2の演算回路22
に送り、前記出力信号と原子炉出力検出器より送付され
た出力信号とから理論給水流量を算出する。
The output signal of the first arithmetic unit 21 is sent to the second arithmetic circuit 22.
The theoretical feed water flow rate is calculated from the output signal and the output signal sent from the reactor power detector.

しかる後に同理論給水流量は補償回路23に送られ、流
量計16から送られてきた実際の給水流量と比較されて
理論給水流量と実給水流量との差すなわち流量差が算出
される。
Thereafter, the theoretical water supply flow rate is sent to the compensation circuit 23, where it is compared with the actual water supply flow rate sent from the flow meter 16, and the difference between the theoretical water supply flow rate and the actual water supply flow rate, that is, the flow rate difference, is calculated.

一方水位検出器20により蒸気発生器4の水位が検出さ
れ、その実水位信号は比較回路24に送られ、ここで予
め定められた標準水位と比較され、実水位と標準水位と
の水位差が算出される。
On the other hand, the water level of the steam generator 4 is detected by the water level detector 20, and the actual water level signal is sent to the comparison circuit 24, where it is compared with a predetermined standard water level, and the water level difference between the actual water level and the standard water level is calculated. be done.

同水位差は補償回路23に送られ、ここで前記した流量
差と、前記水位差とから蒸気発生器4の水位を標準水位
に保つべき調整流量が算出さ札この結果が制御回路25
に送られ、ここで給水弁10の制御信号が発生し、同制
御信号は給水弁10に送られ、同給水弁10は前記調整
流量に見合うように操作される。
The same water level difference is sent to the compensation circuit 23, where the adjustment flow rate to maintain the water level of the steam generator 4 at the standard water level is calculated from the above-mentioned flow rate difference and the water level difference.This result is sent to the control circuit 25.
A control signal for the water supply valve 10 is generated, the control signal is sent to the water supply valve 10, and the water supply valve 10 is operated in accordance with the adjusted flow rate.

本実施例は上記したような構成および作用を有するから
タービン負荷に見合って原子炉出力が上昇あるいは下降
すると直ちに給水流量が調整されて主蒸気流量の変化、
すなわち蒸気発生器4の水位変化にそなえることにより
従来型装置では充分制御できなかった低負荷領域につい
て自動制御が可能であるという利点を有するものである
Since this embodiment has the above-described configuration and operation, when the reactor output increases or decreases in proportion to the turbine load, the feed water flow rate is immediately adjusted and the main steam flow rate changes.
That is, by preparing for changes in the water level of the steam generator 4, it has the advantage that automatic control is possible in the low load region, which could not be adequately controlled with conventional devices.

【図面の簡単な説明】[Brief explanation of drawings]

図は本発明の1実施例を示す系統説明図である。 4・・・蒸気発生器1.17・・・給水温度検出器、1
8・・・主蒸気温度検出器、19・・・原子炉出力検出
器、21・・・第1の演算回路、22・・・第2の演算
回路。
The figure is a system explanatory diagram showing one embodiment of the present invention. 4...Steam generator 1.17...Feed water temperature detector, 1
8... Main steam temperature detector, 19... Reactor output detector, 21... First arithmetic circuit, 22... Second arithmetic circuit.

Claims (1)

【特許請求の範囲】[Claims] 1 蒸気発生器出口側の主蒸気温度検出器と、前記蒸気
発生器入口側の給水温度検出器と、同給水温度検出器の
出力と前記主蒸気温度検出器の出力とから前記蒸気発生
器における伝達熱量を算出する第1の演算回路と、原子
炉出力検出器と、同原子炉出力検出器の出力と前記第1
の演算回路の出力とから前記蒸気発生器へ供給される給
水の流量設定値を算出する第2の演算回路とを有してな
ることを特徴とする蒸気発生器の水位制御装置。
1. A main steam temperature detector on the steam generator outlet side, a feed water temperature detector on the steam generator inlet side, the output of the feed water temperature detector, and the output of the main steam temperature detector in the steam generator. A first arithmetic circuit that calculates the amount of transferred heat, a reactor power detector, an output of the reactor power detector, and the first arithmetic circuit.
A water level control device for a steam generator, comprising: a second arithmetic circuit that calculates a flow rate setting value of feed water supplied to the steam generator from the output of the arithmetic circuit.
JP3883575A 1975-03-31 1975-03-31 Steam generator water level control device Expired JPS5949484B2 (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
JP3883575A JPS5949484B2 (en) 1975-03-31 1975-03-31 Steam generator water level control device

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
JP3883575A JPS5949484B2 (en) 1975-03-31 1975-03-31 Steam generator water level control device

Publications (2)

Publication Number Publication Date
JPS51113001A JPS51113001A (en) 1976-10-05
JPS5949484B2 true JPS5949484B2 (en) 1984-12-03

Family

ID=12536265

Family Applications (1)

Application Number Title Priority Date Filing Date
JP3883575A Expired JPS5949484B2 (en) 1975-03-31 1975-03-31 Steam generator water level control device

Country Status (1)

Country Link
JP (1) JPS5949484B2 (en)

Families Citing this family (2)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPS5531914A (en) * 1978-08-29 1980-03-06 Mitsubishi Atomic Power Ind Feedwater control device of steam generator in atomic power plant
JPS55128706A (en) * 1979-03-28 1980-10-04 Hitachi Ltd Method of controlling water supply to steam drum

Also Published As

Publication number Publication date
JPS51113001A (en) 1976-10-05

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