JPH0574034B2 - - Google Patents
Info
- Publication number
- JPH0574034B2 JPH0574034B2 JP63297733A JP29773388A JPH0574034B2 JP H0574034 B2 JPH0574034 B2 JP H0574034B2 JP 63297733 A JP63297733 A JP 63297733A JP 29773388 A JP29773388 A JP 29773388A JP H0574034 B2 JPH0574034 B2 JP H0574034B2
- Authority
- JP
- Japan
- Prior art keywords
- flow rate
- core
- differential pressure
- reactor
- support plate
- 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 - Fee Related
Links
- 238000000034 method Methods 0.000 claims description 25
- 238000000691 measurement method Methods 0.000 claims description 10
- 230000004907 flux Effects 0.000 claims description 9
- 230000000694 effects Effects 0.000 claims description 7
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 7
- 238000009835 boiling Methods 0.000 claims description 4
- 239000011800 void material Substances 0.000 claims 1
- 238000009530 blood pressure measurement Methods 0.000 description 13
- 238000005259 measurement Methods 0.000 description 7
- 239000002826 coolant Substances 0.000 description 6
- 238000010586 diagram Methods 0.000 description 5
- 230000004043 responsiveness Effects 0.000 description 5
- 238000004364 calculation method Methods 0.000 description 4
- 238000012544 monitoring process Methods 0.000 description 3
- 230000004044 response Effects 0.000 description 3
- 229920006395 saturated elastomer Polymers 0.000 description 3
- 238000012360 testing method Methods 0.000 description 3
- 238000012937 correction Methods 0.000 description 2
- 238000010926 purge Methods 0.000 description 2
- 238000000746 purification Methods 0.000 description 2
- 238000004458 analytical method Methods 0.000 description 1
- 230000008901 benefit Effects 0.000 description 1
- 230000008859 change Effects 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000012806 monitoring device Methods 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000009257 reactivity Effects 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 230000003068 static effect Effects 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)
Description
【発明の詳細な説明】
〔発明の目的〕
（産業上の利用分野）
本発明は原子炉の炉心流量の測定方法に関する
ものである。DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Industrial Application Field) The present invention relates to a method for measuring the core flow rate of a nuclear reactor.
（従来の技術）
沸騰水型原子炉の炉心流量は原子炉反応度制御
の上から極めて重要なパラメータであり、従来は
再循環ポンプによる一次冷却材再循環駆動流とジ
エツトポンプにより炉心流量をつくりだしてこれ
を制御している。このジエツトポンプは静的な流
量素子であり、このデイフエーザ部の差圧を測定
することと、工場試験において各々の流出係数を
求めることで正確な炉心流量野測定方法を得てい
る。この他には炉心支持板部における差圧につい
ても測定しているが、正確なものは期待できず、
従つて前記測定結果の信頼度を高めるために別途
の測定方法を用いて監視ができるようにしてい
る。(Prior technology) The core flow rate of a boiling water reactor is an extremely important parameter from the standpoint of reactor reactivity control. Conventionally, the core flow rate was created using a primary coolant recirculation drive flow using a recirculation pump and a jet pump. This is under control. This jet pump is a static flow rate element, and an accurate method for measuring the core flow field has been obtained by measuring the differential pressure in the diffuser section and determining each discharge coefficient in factory tests. In addition, the differential pressure at the core support plate is also measured, but it cannot be expected to be accurate.
Therefore, in order to increase the reliability of the measurement results, monitoring can be performed using a separate measurement method.
また最近実用化されつつある、原子炉外部に設
置する再循環ポンプを使用しないインターナルポ
ンプを採用した沸騰水型原子炉プラントにおいて
は、従来のジエツトポンプデユフエーザ部差圧測
定に代わるインターナルポンプデツク部差圧測定
を含めたポンプ出入口部差圧を測定する方法等が
知られている。しかしながらこのインターナルポ
ンプ出入口部差圧流量測定方法については、高精
度ではあるがインターナルポンプは動的な機器の
ため複数のインターナルポンプ回転数の入力条件
を必要とする共に、インターナルポンプの個体差
や原子炉内の冷却材の流動状態の影響を考慮しな
くてはならず、このため複雑な演算作業を必要と
しこのためその応答性に難点があり、従つて従来
に比べて高い精度が得られ難かつた。さらに複数
のインターナルポンプの部分台数運転時において
は、停止しているインターナルポンプを介した逆
流が生じるため、より正確な炉心流量を得ること
が困難であつた。なお前記炉心支持板部差圧を測
定する方法は、第３図の特性図に示すように炉心
支持板部差圧と炉心流量の関係が、その炉出力に
よつて炉内のボンド量が変化するため、これによ
り炉心圧損が代わることと、炉心の経年的な圧損
変化があることから、これが差圧測定に影響を及
ぼすため精度の高い炉心流量計測は得られなかつ
た。このため前記したインターナルポンプ出入口
部差圧測定方法にて得られた測定値によつて校正
して使用する必要があつた。但しこの測定方法は
一旦校正済みのものであれば、特に複雑な演算も
ないため比較的応答性は良好であり、殊に過渡的
なインターナルポンプの部分台数運転時に対し
て、炉心の下部格子板位置では略均一な流速とな
ることが実験的に知られているため、演算処理が
不要である点に特長があつた。 In addition, in boiling water reactor plants that employ internal pumps that do not use recirculation pumps that are installed outside the reactor, which have recently been put into practical use, an interface that replaces the conventional jet pump diffuser differential pressure measurement There are known methods of measuring the differential pressure at the pump inlet and outlet, including the measurement of the differential pressure at the pump deck. However, although this method of measuring the differential pressure flow rate at the inlet and outlet of the internal pump is highly accurate, since the internal pump is a dynamic device, it requires input conditions for multiple internal pump rotational speeds. It is necessary to take into account individual differences and the influence of the flow state of the coolant in the reactor, and this requires complex calculation work, which makes the response difficult. It was difficult to obtain. Furthermore, during partial operation of a plurality of internal pumps, backflow occurs through the stopped internal pumps, making it difficult to obtain a more accurate core flow rate. The method for measuring the differential pressure at the core support plate is based on the relationship between the differential pressure at the core support plate and the core flow rate, as shown in the characteristic diagram in Figure 3, and the amount of bond in the reactor changes depending on the reactor power. As a result, the core pressure drop changes and the pressure drop in the core changes over time, which affects the differential pressure measurement, making it impossible to obtain highly accurate core flow rate measurements. For this reason, it was necessary to calibrate the system using the measured values obtained by the above-mentioned internal pump inlet and outlet differential pressure measurement method. However, once this measurement method has been calibrated, it does not require particularly complicated calculations and has relatively good responsiveness. It has been experimentally known that the flow velocity is approximately uniform at the plate position, so the advantage is that no calculation processing is required.
（発明が解決しようとする課題）
一般にインターナルポンプを採用する沸騰水型
原子炉では、主としてインターナルポンプ出入口
部差圧測定方法と、炉心支持板部差圧測定方法と
をプラントシンテム上の要求に対応して、例えば
プラント性能計算には高精度を要求されるため、
インターナルポンプ出入口部差圧測定方法を使用
し、安全保護系機能には応答性の良い炉心支持板
部差圧測定方法を採用していたため、その選択切
替の繁雑と作動結果の信頼性に問題があつた。(Problems to be Solved by the Invention) In a boiling water reactor that generally employs an internal pump, the method for measuring the differential pressure at the inlet and outlet of the internal pump and the method for measuring the differential pressure at the core support plate are mainly performed on the plant system. In response to demands, for example, plant performance calculations require high accuracy.
The differential pressure measurement method at the inlet and outlet of the internal pump was used, and the responsive differential pressure measurement method at the core support plate was used for the safety protection system function, which caused problems in the complexity of switching between selections and the reliability of the operation results. It was hot.
本発明は上記に鑑みてなされたもので、その目
的とするところは、炉心流量測定に際し応答性の
良好な炉心支持板部差圧測定方法による測定に、
高精度が得られる他の複数の測定方法による結果
から校正を行い、より高精度で信頼性を向上した
炉心流量測定方法を提供することにある。 The present invention has been made in view of the above, and its purpose is to provide a method for measuring core support plate differential pressure with good responsiveness when measuring core flow rate.
The purpose of the present invention is to provide a method for measuring core flow rate with higher accuracy and improved reliability by performing calibration from the results of multiple other measurement methods that provide high accuracy.
（課題を解決するための手段）
炉心支持板部差圧測定手段に、インターナルポ
ンプ出入口部差圧測定手段と炉心下部の入口温度
から原子炉の熱平衡による流量測定手段による校
正及び原子炉内に設けた中性子束検出器による補
正手段を具備する。(Means for solving the problem) Calibration is performed using the core support plate differential pressure measuring means, the internal pump inlet and outlet differential pressure measuring means, and the flow rate measuring means based on the thermal balance of the reactor based on the inlet temperature at the lower part of the core. A correction means is provided using a neutron flux detector provided.
（作用）
応答性が良好であるが冷却材のボイド量の影響
を受け易い、炉心支持板部差圧測定方法による炉
心流量の測定値を、測定精度は高いが運転ポンプ
の台数等各種条件の考慮が必要なインターナルポ
ンプ出入口部差圧測定による流量測定値及び原子
炉の熱平衡演算による流量測定値により定期的に
校正すると共に、中性子束検出器により検出した
冷却材のボイド量による影響を補正して、高精度
で応答性良く炉心流量の測定を行う。(Function) The core flow rate measured by the core support plate differential pressure measurement method, which has good response but is easily affected by the amount of voids in the coolant, has high measurement accuracy but is sensitive to various conditions such as the number of operating pumps. Regularly calibrate the flow rate measured by measuring the differential pressure at the entrance and exit of the internal pump, which must be taken into consideration, and the flow rate measured by calculating the thermal balance of the reactor, and also correct the influence of the amount of voids in the coolant detected by the neutron flux detector. The reactor core flow rate is measured with high precision and responsiveness.
（実施例）
本発明の一実施例を図面を参照して説明する。
第１図は全体構成図で、原子炉圧力容器１内で炉
心２の周囲下部に設置された複数のインターナル
ポンプ３により、原子炉圧力容器１内の冷却材４
が炉心２の下部より上部に向かう流れがつくられ
る。また炉心２の下部には炉心支持板５があり、
これを挟んで上下部に開口して設けたノズルに接
続した計装配管６ａ，６ｂを介して連結した支持
板部差圧伝送器７を設け、この出力の差圧信号は
支持板部流量演算器８に入力する。また前記イン
ターナルポンプ３のポンプデツク９部あるいはイ
ンターナルポンプ３の出入口部に開口して設けた
ノズルより計装配管１０ａ，１０ｂを介して連結
した出入口部差圧伝送器１１を設け、この出力の
差圧信号は出入口部流量演算器１２に入力され
る。さらに前記炉心２内には中性子束検出器１３
と、炉心２の下部入口に冷却材４の温度検出器１
４を設置して、中性子束検出器１３からの中性子
束信号は平均出力領域回路１５を介して前記支持
板部流量演算器８に入力する。温度検出器１４か
らの温度信号は熱平衡流量演算器１６に入力され
る。この熱平衡流量演算器１６では、この際ダウ
ンカマ１６及び炉心支持板５の入口部の温度を前
記温度検出器１４で代表させて、原子炉へのエネ
ルギー熱収支を演算し、この原子炉の熱平衡（ヒ
ートバランス）測定方法により炉心流量値を求め
て、この出力を前記支持板部流量演算器８に出力
する。また出入口部流量演算器１２においては、
インターナルポンプ３の出入口部からの出入口部
差圧信号と、図示しない各インターナルポンプ３
の回転数情報及び各インターナルポンプ３の工場
における試験結果から得られたＱ＝Ｈ特性情報か
ら、インターナルポンプ出入口部差圧測定方法に
よる炉心流量を算出し、支持板部流量演算器８に
出力すると共に、必要に応じて他の用途として指
示計または記録計１７に出力する。前記支持板部
流量演算器８においては、炉心支持板部差圧測定
方法により炉心流量を算出して、その出力はシス
テムの要求に従いこの結果を指示計または記録計
１８に出力すると共に、流量制御信号や他の機器
に対するインターロツク信号とするように構成さ
れている。(Example) An example of the present invention will be described with reference to the drawings.
FIG. 1 is an overall configuration diagram, in which a plurality of internal pumps 3 installed around the lower part of the reactor core 2 in the reactor pressure vessel 1 pump the coolant inside the reactor pressure vessel 1.
A flow is created from the bottom of the core 2 toward the top. In addition, there is a core support plate 5 at the bottom of the core 2.
A support plate differential pressure transmitter 7 is provided which is connected via instrumentation pipes 6a and 6b connected to nozzles opened at the top and bottom of this, and the output differential pressure signal is used to calculate the support plate flow rate. input into device 8. In addition, an inlet/outlet differential pressure transmitter 11 is provided which is connected to the pump deck 9 of the internal pump 3 or a nozzle opened at the inlet/outlet of the internal pump 3 via instrumentation piping 10a, 10b. The differential pressure signal is input to the inlet/outlet flow rate calculator 12 . Furthermore, a neutron flux detector 13 is located inside the core 2.
and a temperature sensor 1 for the coolant 4 at the lower inlet of the core 2.
4 is installed, and the neutron flux signal from the neutron flux detector 13 is inputted to the support plate flow rate calculator 8 via the average output area circuit 15. A temperature signal from the temperature detector 14 is input to a thermal equilibrium flow rate calculator 16 . The thermal equilibrium flow rate calculator 16 uses the temperature detector 14 to represent the temperature at the inlet of the downcomer 16 and the core support plate 5, calculates the energy heat balance to the reactor, and calculates the thermal balance of the reactor ( The core flow rate value is determined by the heat balance) measurement method, and this output is output to the support plate flow rate calculator 8. In addition, in the inlet/outlet flow rate calculator 12,
Inlet/outlet differential pressure signals from the inlet/outlet of the internal pump 3 and each internal pump 3 (not shown)
From the rotational speed information and the Q=H characteristic information obtained from the factory test results of each internal pump 3, the core flow rate is calculated by the internal pump inlet and outlet differential pressure measurement method, and the flow rate is calculated by the support plate flow rate calculator 8. At the same time, it is also output to an indicator or recorder 17 for other purposes as necessary. The support plate flow rate calculator 8 calculates the core flow rate by the core support plate differential pressure measurement method, and outputs the result to the indicator or recorder 18 according to system requirements, and also controls the flow rate. It is configured to serve as an interlock signal for signals and other equipment.
次に上記構成による作用について説明する。第
２図は作動フロー図で、支持板部流量演算器８に
おいて、炉心支持板部差圧（ΔP）から炉心支持
板部差圧測定方法により応答性の良い炉心流量を
算出するが、この測定方法では原子炉の出力に応
じ発生するボイド量の影響から支持板部差圧伝送
器７からの差圧信号に変化が生じるため、中性子
束信号から平均出力領域回路１５を介して入力さ
れた出力補正信号（Ｐ）より、差圧変化分を予め
解析や原子炉の起動試験にて求めてある炉心流量
−炉心支持板部差圧特性により常時補正を行な
う。またこの支持板部流量演算器８は前記出入口
部流量演算器１２からのインターナルポンプ出入
口部差圧測定方法による炉心流量信号により定期
的に校正すると共に、さらに前記熱平衡流量演算
器１６からの熱平衡測定方法による炉心流量信号
（Ｃ）によつても定期的に校正して精度を向上さ
せた炉心流量信号（Ｗ）を演算出力する。この誤
差の少ない炉心流量信号は従来と同様に炉心流量
測制御のほか、例えば中性子束監視上の基準流量
測定や炉心流量の急速低下に対するスクラム機能
等、保護系の各種インターロツクや監視装置に伝
達されて原子炉運転の信頼性を向上する。 Next, the effect of the above configuration will be explained. Figure 2 is an operational flow diagram, in which the support plate flow rate calculator 8 calculates the core flow rate with good responsiveness from the core support plate differential pressure (ΔP) using the core support plate differential pressure measurement method. In this method, since the differential pressure signal from the support plate differential pressure transmitter 7 changes due to the effect of the amount of voids generated depending on the output of the reactor, the output input from the neutron flux signal via the average power region circuit 15 changes. Based on the correction signal (P), the differential pressure change is constantly corrected based on the core flow rate-core support plate differential pressure characteristic obtained in advance through analysis or reactor start-up tests. The support plate flow rate calculator 8 is periodically calibrated using the core flow rate signal from the inlet/outlet flow rate calculator 12 using the internal pump inlet/outlet differential pressure measurement method, and furthermore, the support plate flow rate calculator 8 is calibrated using the core flow rate signal from the inlet/outlet flow rate calculator 12 using the internal pump inlet/outlet differential pressure measuring method. The core flow rate signal (C) determined by the measurement method is also periodically calibrated to calculate and output a core flow rate signal (W) with improved accuracy. This core flow rate signal with less error is transmitted to various interlocks and monitoring devices in the protection system, such as the reference flow rate measurement for neutron flux monitoring and the scram function for rapid decreases in core flow rate, in addition to the core flow measurement control as in the past. This will improve the reliability of reactor operation.
ここで熱平衡（ヒートバランス）測定方法によ
る炉心流量Wtの概略計算は、下記の(1)式による
ことができる。 Here, the core flow rate W t can be roughly calculated by the following equation (1) using the heat balance measurement method.
Wt＝Wcr・（hf−hcr）＋Wp（hf−hp）＋Wcuw・（hf−hc
uw）＋Wfw（hf−hfw）＋C1・（QL−QP）／hf−hp＋fcu
・hfg……(1)
ここで、
hp；炉心入口エンタルピ、
hf；飽和水エンタルピ、
hfg；飽和蒸気エンタルピ、
hcu；ダウンカマに入る飽和蒸気流量、
Wcr；制御棒駆動系からの流量、
hcr；制御棒駆動系からのエンタルピ、
Wp；インターナルポンプパージ流量、
hp；インターナルポンプパージエンタルピ、
Wcuw；原子炉浄化系からの流入流量、
hcuw；原子炉浄化系からの流入エンタルピ、
Wfw；原子炉給水流量、
hfw；原子炉給水エンタルピ、
QL；ダウンカマ部で失われる熱エネルギー、
C1；変換定数（860Kcal／kWh）、
QL；インターナルポンプによつて得られる熱
エネルギー、である。W t = W cr・(h f −h cr )＋W p (h f −h p )＋W cuw・(h f −h c
uw )＋W fw (h f −h fw )＋C 1・(Q L −Q P )/h f −h p +f cu
・h fg ……(1) where, h p ; core inlet enthalpy, h f : saturated water enthalpy, h fg : saturated steam enthalpy, h cu : saturated steam flow rate entering the downcomer, W cr : from control rod drive system flow rate, h cr ; enthalpy from control rod drive system, W p ; internal pump purge flow rate, h p ; internal pump purge enthalpy, W cuw : inflow flow rate from reactor purification system, h cuw ; reactor purification flow rate. Inflow enthalpy from the system, W fw ; Reactor feed water flow rate, h fw ; Reactor feed water enthalpy, Q L : Thermal energy lost in the downcomer section, C 1 : Conversion constant (860Kcal/kWh), Q L : Internal pump The thermal energy obtained by
従つて炉心支持板部差圧測定方法単独の場合に
生ずる測定誤差（100％）に対して、先ずインタ
ーナルポンプ出入口部差圧測定方法による誤差校
正の効果は下記(2)式のようになる。 Therefore, for the measurement error (100%) that occurs when only the core support plate differential pressure measurement method is used, the effect of error calibration using the internal pump inlet and outlet differential pressure measurement method is as shown in equation (2) below. .
Σ1＝√a 2＋c 2 ……(2)
ここで、ea；ポンプ出入口部差圧測定上の誤
差、
ec；炉心支持板部差圧測定上の誤差。 Σ 1 =√ a 2 + c 2 ...(2) where, e a ; Error in measuring the differential pressure at the pump inlet and outlet; e c ; Error in measuring the differential pressure at the core support plate.
またに熱平衡よりの流量測定方法による誤差校
正の効果は下記(3)式のようになる。 Furthermore, the effect of error calibration using the flow rate measurement method based on thermal equilibrium is expressed by equation (3) below.
Σ2＝√１４（a 2＋b 2）＋c 2 ……(3) ここで、eb；熱平衡より測定する誤差。 Σ 2 = √14 ( a 2 + b 2 ) + c 2 ...(3) where, e b ; error measured from thermal equilibrium.
以上から、いまea＝eb＝ecと仮定すると、本発
明によりΣ2／Σ1＝√３／２＝87％に縮小改善す
ることができる。 From the above, assuming that e a = e b = e c , the present invention can improve the reduction to Σ 2 /Σ 1 =√3/2=87%.
以上本発明によれば、インターナルポンプ採用
の原子炉の炉心流量測定に際し、炉心支持板部差
圧測定方法の結果を他の異なる測定方法による測
定値により定期的、連続的に校正及び補正をする
ことにより、複数のインターナルポンプの種々の
運転状態においも、応答性が良好で、しかも正確
な炉心流量が得られるため原子炉の出力制御、運
転監視、保護等が高精度に行われ、原子炉運転の
信頼性を向上する効果がある。
As described above, according to the present invention, when measuring the core flow rate of a nuclear reactor that employs an internal pump, the results of the core support plate differential pressure measurement method are periodically and continuously calibrated and corrected using the values measured by other different measurement methods. By doing so, even in various operating conditions of multiple internal pumps, responsiveness is good and accurate core flow rate can be obtained, so reactor output control, operation monitoring, protection, etc. can be performed with high precision. This has the effect of improving the reliability of nuclear reactor operation.
第１図は本発明の一実施例の全体構成図、第２
図は本発明による動作フロー図、第３図は炉出力
の変化に対する炉心流量と炉心支持板部差圧の特
性図である。
２……炉心、３……インターナルポンプ、５…
…炉心支持板、７……炉心支持板部差圧伝送器、
８……支持板部流量演算器、１１……出入口部差
圧伝送器、１２……入口部流量演算器、１３……
中性子束検出器、１４……温度検出器、１５……
平均出力領域回路、１６……熱平衡流量演算器、
１７，１８……記録計。
Fig. 1 is an overall configuration diagram of an embodiment of the present invention, Fig. 2
The figure is an operation flowchart according to the present invention, and FIG. 3 is a characteristic diagram of core flow rate and core support plate differential pressure with respect to changes in reactor output. 2...Reactor core, 3...Internal pump, 5...
...Core support plate, 7...Core support plate differential pressure transmitter,
8...Support plate part flow rate calculator, 11...Inlet/outlet part differential pressure transmitter, 12...Inlet part flow rate calculator, 13...
Neutron flux detector, 14... Temperature detector, 15...
Average output area circuit, 16... thermal equilibrium flow rate calculator,
17, 18...Recorder.
Claims (1)
ーナルポンプを設置してなる沸騰水型原子炉にお
いて、炉心支持板部差圧による流量測定方法に対
して、インターナルポンプの出入口部差圧の流量
測定方法及び炉心下部入口温度からの原子炉の熱
平衡による流量測定方法により算出した炉心流量
測定値により校正を行うと共に、炉内中性子束信
号により炉心圧損のボイド量の影響による変化分
を補正することを特徴とする炉心流量測定方法。1. In a boiling water reactor in which an internal pump is installed in the reactor to control the core flow rate, the method of measuring the flow rate using the differential pressure at the core support plate is different from the method using the differential pressure at the entrance and exit of the internal pump. Calibration is performed using the measured value of the core flow rate calculated by the flow rate measurement method and the flow rate measurement method based on the thermal balance of the reactor from the lower core inlet temperature, and the in-core neutron flux signal is used to correct for changes in core pressure drop due to the effect of void volume. A method for measuring reactor core flow rate, characterized by:
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63297733A JPH02143194A (en) | 1988-11-25 | 1988-11-25 | Flow rate measuring method in core |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63297733A JPH02143194A (en) | 1988-11-25 | 1988-11-25 | Flow rate measuring method in core |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH02143194A JPH02143194A (en) | 1990-06-01 |
| JPH0574034B2 true JPH0574034B2 (en) | 1993-10-15 |
Family
ID=17850475
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP63297733A Granted JPH02143194A (en) | 1988-11-25 | 1988-11-25 | Flow rate measuring method in core |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH02143194A (en) |
-
1988
- 1988-11-25 JP JP63297733A patent/JPH02143194A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPH02143194A (en) | 1990-06-01 |
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