JP2001141874A - Flow rate-measuring device for reactor core cooling material - Google Patents
Flow rate-measuring device for reactor core cooling materialInfo
- Publication number
- JP2001141874A JP2001141874A JP32503399A JP32503399A JP2001141874A JP 2001141874 A JP2001141874 A JP 2001141874A JP 32503399 A JP32503399 A JP 32503399A JP 32503399 A JP32503399 A JP 32503399A JP 2001141874 A JP2001141874 A JP 2001141874A
- Authority
- JP
- Japan
- Prior art keywords
- flow rate
- core
- coolant flow
- core coolant
- reactor
- Prior art date
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Classifications
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- 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
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- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
【０００１】[0001]
【発明の属する技術分野】本発明は再循環ポンプ内蔵型
沸騰水型原子炉の炉心冷却材の測定法に関するものであ
る。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for measuring core coolant of a boiling water reactor with a built-in recirculation pump.
【０００２】[0002]
【従来の技術】一般に、沸騰水型原子炉では核***によ
って発生した熱を水で冷却しているが、冷却水は熱を除
去するだけではなく、核***によって発生した高速中性
子を熱中性子に減速する働きがある。これは中性子とウ
ラン燃料の核***反応を起きやすくするためである。一
方、炉心冷却材流量によって炉心の冷却水と蒸気の割合
が変化することから、流量の変化によって炉心の出力も
変化することになる。したがって、炉心冷却水の流量を
精度よく測定することは、原子炉の運転監視において非
常に重要である。2. Description of the Related Art In a boiling water reactor, heat generated by nuclear fission is generally cooled by water, but the cooling water not only removes heat but also decelerates fast neutrons generated by nuclear fission to thermal neutrons. There is work. This is to facilitate the fission reaction between neutron and uranium fuel. On the other hand, since the ratio of cooling water and steam in the core changes according to the flow rate of the core coolant, the output of the core also changes according to the change in the flow rate. Therefore, it is very important to accurately measure the flow rate of the core cooling water in the operation monitoring of the reactor.
【０００３】米国特許4、975、239号明細書に、原子炉
内の冷却材流量を測定する方法として、支持板差圧測定
法が示されている。この測定法では熱水力解析手法を用
いて、いくつかの炉心出力／炉心冷却材流量点における
支持板差圧の予測値を求め、そこから炉心出力と支持板
差圧より炉心冷却材流量が算出される相関式を予め作成
する。そして、測定された炉心熱出力と支持板の上下の
差圧から炉心冷却材流量を推定している。[0003] US Pat. No. 4,975,239 discloses a method of measuring a differential pressure of a support plate as a method of measuring a coolant flow rate in a nuclear reactor. In this measurement method, the predicted value of the support plate differential pressure at several core power / core coolant flow rate points was calculated using a thermal hydraulic analysis method, and the core coolant flow rate was calculated from the core power and the support plate differential pressure based on this. The calculated correlation equation is created in advance. Then, the core coolant flow rate is estimated from the measured core heat output and the differential pressure above and below the support plate.
【０００４】また、内蔵型の再循環ポンプ（以下インタ
ーナルポンプと呼ぶ）の入口と出口の差圧を測定し、予
め求めたQ−H特性を用いて炉心の冷却材流量を推定する
ポンプデッキ部差圧法が一般に知られている。Further, a pump deck for measuring a differential pressure between an inlet and an outlet of a built-in recirculation pump (hereinafter referred to as an internal pump) and estimating a coolant flow rate in a core using a QH characteristic obtained in advance. The partial pressure difference method is generally known.
【０００５】特開平9−122782号公報にはポンプデッキ
部差圧測定法と支持板差圧測定法において、各々熱収支
法との繰り返し計算により、炉心入口流量および冷却材
温度を正確に求め、評価する方法が記述されている。Japanese Patent Application Laid-Open No. Hei 9-122782 discloses that in a pump deck part differential pressure measuring method and a support plate differential pressure measuring method, a core inlet flow rate and a coolant temperature are accurately obtained by repeatedly calculating each heat balance method. The method of evaluation is described.
【０００６】[0006]
【発明が解決しようとする課題】しかしながら、前述の
ようにポンプ部差圧法は、インターナルポンプのＱ−Ｈ
特性から冷却材流量を推定するのであるが、Ｑ−Ｈ特性
として工場試験結果が必要であり、プラント寿命中のポ
ンプ交換等を行なう場合、再設定を要すると考えられ、
ポンプによらない合理的な測定手法が望まれる。However, as described above, the pump differential pressure method uses the QH of the internal pump.
Although the coolant flow rate is estimated from the characteristics, the factory test results are required as the QH characteristics, and it is considered that resetting is required when performing pump replacement during the life of the plant.
A rational measurement method not using a pump is desired.
【０００７】前記米国発明の炉心支持板差圧測定法は、
この課題に対する方策の一つと考えられるが、出力と炉
心差圧以外の影響を考慮できない。[0007] The method for measuring the core support plate differential pressure of the United States invention is as follows.
Although this is considered to be one of the measures to solve this problem, it is impossible to consider effects other than power and core pressure difference.
【０００８】本発明の目的は、変化時間の長い不確定要
因の影響を考慮して、沸騰水型原子炉の冷却材炉心流量
を精度良く推定することができる炉心支持板差圧測定法
を提供することである。An object of the present invention is to provide a method for measuring a core support plate differential pressure capable of accurately estimating a coolant core flow rate of a boiling water reactor in consideration of the influence of an uncertain factor having a long change time. It is to be.
【０００９】[0009]
【課題を解決するための手段】本発明の炉心冷却材流量
計測方法及び炉心冷却材流量計測装置は、上記課題を解
決するため、本発明の請求項１においては、炉心性能計
算機に炉心冷却材流量演算器、長期トレンド流量演算器
及び流量比較演算器を備え、炉心冷却材流量演算器で算
出した炉心冷却材流量と長期トレンド流量演算器で算出
される長期トレンド流量の差がある値より小さくなるよ
うに、炉心冷却材流量計算で用いる入力定数を調整する
ことにより、変化時間の長い不確定要因の影響による炉
心冷却材流量測定誤差を補正する手段を有することを特
徴とするものである。In order to solve the above-mentioned problems, a core coolant flow rate measuring method and a core coolant flow rate measuring apparatus according to the present invention are described in claim 1 of the present invention. Equipped with a flow rate calculator, long-term trend flow rate calculator and flow rate comparison calculator, the difference between the core coolant flow rate calculated by the core coolant flow rate calculator and the long-term trend flow rate calculated by the long-term trend flow rate calculator is smaller than a certain value. The present invention is characterized by having a means for correcting a core coolant flow rate measurement error due to the influence of an uncertain factor having a long change time by adjusting an input constant used in core coolant flow rate calculation.
【００１０】また、本発明の請求項２においては、前記
の炉心冷却材流量演算器で行われる炉心冷却材流量計算
において、測定された炉心支持板差圧、原子炉内のヒー
トバランス及び炉心軸方向出力分布より炉心冷却材流量
を推定する手段を有することを特徴とするものである。According to a second aspect of the present invention, in the core coolant flow rate calculation performed by the core coolant flow rate calculator, the measured core support plate differential pressure, the heat balance in the reactor, and the core axis. It is characterized by having means for estimating the core coolant flow rate from the directional power distribution.
【００１１】さらに、本発明の請求項３においては、前
記の炉心冷却材流量計算において、燃料表面のクラッド
厚さに対応した入力定数を調整し、炉心冷却材流量を補
正することを特徴とするものである。Further, according to a third aspect of the present invention, in the core coolant flow rate calculation, an input constant corresponding to a clad thickness on a fuel surface is adjusted to correct the core coolant flow rate. Things.
【００１２】そして、本発明の請求項４においては、前
記の炉心冷却材流量計算において、燃料表面の摩擦係数
を調整し、炉心冷却材流量を補正することを特徴とする
ものである。According to a fourth aspect of the present invention, in the calculation of the core coolant flow rate, the friction coefficient of the fuel surface is adjusted to correct the core coolant flow rate.
【００１３】また、本発明の請求項５においては、前記
の炉心冷却材流量計算において、燃料の表面粗さに対応
する入力定数を調整し、炉心冷却材流量を補正すること
を特徴とするものである本発明の請求項６においては、
前記の長期トレンド流量演算器において、原子炉周りの
熱収支から算出するため、燃料へのクラッド付着等の変
化時間の長い不確定要因の影響を受けない、ヒートバラ
ンス流量にフィルター処理を行った、長期トレンド流量
を算出することを特徴とするものである。According to a fifth aspect of the present invention, in the core coolant flow rate calculation, an input constant corresponding to the surface roughness of the fuel is adjusted to correct the core coolant flow rate. In claim 6 of the present invention,
In the long-term trend flow calculator, to calculate from the heat balance around the reactor, unaffected by a long uncertain factor of change time such as clad adhesion to the fuel, the heat balance flow rate was filtered, It is characterized by calculating a long-term trend flow rate.
【００１４】[0014]
【発明の実施の形態】以下、本発明の第1の実施例を図
１及び図２を参照して説明する。図１に本発明を実施す
るための機器構成全体を示す。原子炉圧力計１は圧力容
器２内の圧力を測定し、温度計３はダウンカマ部の冷却
水温度を測定している。これらより炉心入口エンタルピ
ーを算出する。また、差圧計４でポンプ部の入口と出口
の差圧を測定し、差圧計５を用いて炉心支持板の上下の
差圧を測定する。さらに、流量計、温度計及び圧力計６
を用いて給水流量、給水の温度及びそのときの圧力を測
定する。また、中性子検出器７を用いて圧力容器内の局
所的な中性子量を測定し、炉心の軸方向出力分布を算出
する。これらの測定値を炉心性能計算機９に取り込み、
炉心冷却材流量演算器１０、長期トレンド流量演算器１
１及び比較演算器１２を用いて炉心冷却材流量を決定す
る。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Hereinafter, a first embodiment of the present invention will be described with reference to FIGS. FIG. 1 shows the entire device configuration for implementing the present invention. The reactor pressure gauge 1 measures the pressure in the pressure vessel 2, and the thermometer 3 measures the cooling water temperature in the downcomer part. From these, the core enthalpy is calculated. The differential pressure gauge 4 measures the differential pressure between the inlet and the outlet of the pump unit, and the differential pressure gauge 5 measures the differential pressure between the upper and lower parts of the core support plate. Furthermore, a flow meter, a thermometer and a pressure gauge 6
Is used to measure the feedwater flow rate, feedwater temperature and pressure at that time. Further, the neutron detector 7 is used to measure the local neutron amount in the pressure vessel, and calculate the axial power distribution of the core. These measured values are taken into the core performance calculator 9,
Core coolant flow rate calculator 10, long-term trend flow rate calculator 1
1 and the comparator 12 are used to determine the core coolant flow rate.
【００１５】次に本測定法の第１の実施例を図２に示す
フローチャートを用いて説明する。Next, a first embodiment of the present measuring method will be described with reference to a flowchart shown in FIG.
【００１６】ステップ１及びステップ２は炉心冷却材流
量演算器において行われる。ステップ１では、測定によ
って得られる入口エンタルピーを用いて、ある炉心冷却
材流量に対する熱出力を（数１）で表される原子炉内の
熱収支の関係式から計算する。Steps 1 and 2 are performed in the core coolant flow rate calculator. In Step 1, using the inlet enthalpy obtained by the measurement, the heat output with respect to a certain core coolant flow rate is calculated from the relational expression of the heat balance in the reactor represented by (Equation 1).
【００１７】[0017]
【数１】 (Equation 1)
【００１８】次にステップ２の炉心冷却材流量計算にお
いて、測定によって得られる炉心軸方向出力分布、圧力
容器内圧力、入口エンタルピー及びステップ１で算出さ
れる熱出力から、炉心冷却材流量を推定する。ステップ
１〜２の計算を繰り返し行なうことにより、原子炉内の
ヒートバランスと炉心冷却材流量計算を同時に満足する
熱出力と炉心冷却材流量の収束解を求める。Next, in the calculation of the core coolant flow rate in step 2, the core coolant flow rate is estimated from the core axial power distribution obtained by the measurement, the pressure in the pressure vessel, the inlet enthalpy, and the heat output calculated in step 1. . By repeatedly performing the calculations in steps 1 and 2, a convergence solution of the heat output and the core coolant flow rate that simultaneously satisfies the heat balance in the reactor and the core coolant flow rate calculation is obtained.
【００１９】一方、ステップ３において、原子炉周りも
含めた熱収支から算出されるヒートバランス流量に、１
次遅れのフィルター処理を行なう。ヒートバランス流量
は給水流量、給水温度等多くのプロセス量の測定に基づ
くため、指示値のバラツキが大きいが、長期的なポンプ
のQ−H特性の変動や燃料に付着したクラッド等による長
期的な炉心圧損の変動の影響を受けない。そこでヒート
バランス流量にフィルター処理を行なうことによって、
バラツキを小さくし、炉心冷却材流量の瞬時の変動を示
すことはできないが、炉心冷却材流量の長期的なトレン
ド変化をよく示す長期トレンド流量に変換する。On the other hand, in Step 3, the heat balance flow rate calculated from the heat balance including around the reactor
Next-order filter processing is performed. Since the heat balance flow rate is based on the measurement of many process quantities such as feed water flow rate and feed water temperature, there are large variations in the indicated values, but long-term fluctuations in Q-H characteristics of the pump and long-term Unaffected by fluctuations in core pressure loss. So by filtering the heat balance flow rate,
Although the variation cannot be reduced and the instantaneous fluctuation of the core coolant flow rate cannot be shown, it is converted into a long-term trend flow rate that well indicates a long-term trend change of the core coolant flow rate.
【００２０】さらにステップ４では、ステップ１、２で
求めた炉心冷却材流量とステップ３で求めた長期トレン
ド流量を比較する。このとき、その差がある値より大き
ければ、ステップ５において炉心冷却材流量計算で用い
ている適当な入力定数を変更し、改めて熱出力及び炉心
冷却材流量の収束解を算出する。そして、長期トレンド
流量との差がある値より小さくなるまでこの操作を繰り
返し、最終的な炉心冷却材流量を決定する。ここで入力
定数として具体的には燃料クラッド厚さを調整する。In step 4, the core coolant flow rate obtained in steps 1 and 2 is compared with the long-term trend flow rate obtained in step 3. At this time, if the difference is larger than a certain value, an appropriate input constant used in the core coolant flow rate calculation is changed in step 5, and a convergence solution of the heat output and the core coolant flow rate is calculated again. This operation is repeated until the difference from the long-term trend flow becomes smaller than a certain value, and the final core coolant flow is determined. Here, the fuel constant is specifically adjusted as the input constant.
【００２１】燃料クラッド厚さを調整すると燃料集合体
の流路面積が変化し、燃料集合体の圧損が変化する。実
機燃料において燃料クラッド厚さを運転中に実測するこ
とは困難であるが、出力分布変化変化以外の圧損変化の
要因として考えられるため、本実施例ではクラッド厚さ
を調整することとしている。When the thickness of the fuel clad is adjusted, the flow passage area of the fuel assembly changes, and the pressure loss of the fuel assembly changes. Although it is difficult to measure the fuel clad thickness of the actual fuel during operation, it is considered as a factor of a pressure loss change other than a change in the power distribution. Therefore, in the present embodiment, the clad thickness is adjusted.
【００２２】また、起動試験の時点で、ポンプデッキ部
差圧測定法を用いて算出される炉心冷却材流量を絶対校
正流量として、炉心冷却材流量演算器や長期トレンド流
量演算器で計算される各流量の補正を行なうことによ
り、より正確な流量測定を行なうことができる。At the time of the start-up test, the core coolant flow rate calculated using the pump deck differential pressure measuring method is calculated as an absolute calibration flow rate by a core coolant flow rate calculator or a long-term trend flow rate calculator. By correcting each flow rate, more accurate flow rate measurement can be performed.
【００２３】図３に実際のプラントデータから本発明の
第１の実施例を用いて求めた炉心冷却材流量推定値と本
発明を適用していない炉心冷却材流量推定値及び長期ト
レンド流量（ヒートバランス流量にフィルター処理した
もの）を示す。本発明を適用していない炉心冷却材流量
推定値はある時点から長期トレンド流量と大きくずれが
生じているのに対して、本発明を適用した炉心冷却材流
量推定値は、測定誤差が補正されている。FIG. 3 shows an estimated value of the core coolant flow rate obtained from the actual plant data using the first embodiment of the present invention, an estimated value of the core coolant flow rate to which the present invention is not applied, and a long-term trend flow rate (heat (Filtered to balance flow rate). While the estimated value of the core coolant flow rate to which the present invention is not applied has a large deviation from the long-term trend flow rate from a certain point in time, the estimated value of the core coolant flow rate to which the present invention is applied has a measurement error corrected. ing.
【００２４】図４に本発明の第２の実施例を示す。本実
施例は、第１の実施例のステップ５の入力定数として燃
料棒表面の摩擦係数または表面粗さを調整するものであ
る。燃料摩擦係数を調整すると燃料集合体表面に付着し
たクラッド等による摩擦圧損係数の変化に対応したもの
といえる。実機燃料において燃料摩擦圧損係数や表面粗
さ運転中に実測することは困難であるが、出力分布変化
やクラッド厚さの変化以外の圧損変化の要因として考え
られるため、本実施例では燃料摩擦圧損係数または表面
粗さを調整することとしている。FIG. 4 shows a second embodiment of the present invention. In this embodiment, the friction coefficient or the surface roughness of the fuel rod surface is adjusted as an input constant in step 5 of the first embodiment. It can be said that adjusting the fuel friction coefficient corresponds to a change in the friction pressure loss coefficient due to the clad or the like attached to the fuel assembly surface. Although it is difficult to measure the fuel friction pressure loss coefficient and surface roughness of actual fuel during operation, it is considered to be a factor of pressure loss change other than a change in power distribution and a change in clad thickness. The coefficient or surface roughness is to be adjusted.
【００２５】[0025]
【発明の効果】以上述べたように、本発明の請求項１に
よれば、沸騰水型原子炉において、出力分布やクラッド
付着による炉心状態の変化の影響による炉心冷却材流量
測定誤差を補正し、正確で精度の高い冷却材炉心流量計
測を行うことができる効果がある。As described above, according to the first aspect of the present invention, in the boiling water reactor, the core coolant flow rate measurement error due to the power distribution and the influence of the core state change due to the clad adhesion is corrected. Thus, there is an effect that the coolant core flow rate measurement can be performed accurately and accurately.
【図１】本発明の第１の実施例の構成図。FIG. 1 is a configuration diagram of a first embodiment of the present invention.
【図２】本発明の第１の実施例のフローチャート。FIG. 2 is a flowchart of the first embodiment of the present invention.
【図３】本発明を適用することによって求められる炉心
冷却材流量の時間変化を示す図。FIG. 3 is a diagram showing a temporal change of a core coolant flow rate obtained by applying the present invention.
【図４】本発明の第２の実施例のフローチャート。FIG. 4 is a flowchart of a second embodiment of the present invention.
１…原子炉圧力容器内圧力計、２…原子炉圧力容器、３
…炉心入口部温度計、４…ポンプデッキ部差圧計、５…
炉心支持板差圧計、６…給水流量計・給水温度計・給水
圧力計、７…中性子検出器、９…炉心性能計算機、１０
…炉心冷却材流量演算器、１１…長期トレンド流量演算
器、１２…流量比較演算器。1: pressure gauge in reactor pressure vessel, 2: reactor pressure vessel, 3
... thermometer at core inlet, 4 ... differential pressure gauge at pump deck, 5 ...
Core support plate differential pressure gauge, 6: feed water flow meter, feed water temperature gauge, feed water pressure gauge, 7: neutron detector, 9: core performance calculator, 10
... Core coolant flow rate calculator, 11 ... Long term trend flow rate calculator, 12 ... Flow rate comparison calculator.
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Claims (6)
量演算器、長期トレンド流量演算器及び流量比較演算器
を備え、炉心冷却材流量演算器で算出した炉心冷却材流
量と長期トレンド流量演算器で算出される長期トレンド
流量の差がある値より小さくなるように炉心冷却材流量
計算の入力定数を調整することにより、炉心冷却材流量
を補正することを特徴とする原子炉炉心流量計算方法及
び炉心冷却材流量測定装置。1. A boiling water reactor comprising a core coolant flow rate calculator, a long-term trend flow rate calculator, and a flow rate comparison calculator, wherein a core coolant flow rate calculated by the core coolant flow rate calculator and a long-term trend flow rate calculation are provided. Core flow rate is corrected by adjusting the input constant of the core flow rate calculation so that the difference in the long-term trend flow rate calculated by the reactor becomes smaller than a certain value. And core coolant flow rate measuring device.
却材流量計算において、測定された炉心支持板差圧、原
子炉内のヒートバランス及び炉心軸方向出力分布より炉
心冷却材流量を推定することを特徴とする請求項１記載
の原子炉炉心流量計算方法及び炉心冷却材流量測定装
置。2. In a core coolant flow rate calculation performed by a core coolant flow rate calculator, a core coolant flow rate is estimated from a measured core support plate differential pressure, a heat balance in a reactor, and a core axial power distribution. The reactor core flow rate calculation method and the core coolant flow rate measuring apparatus according to claim 1, characterized in that:
のクラッド厚さに対応した入力定数を調整し、炉心冷却
材流量を補正することを特徴とする請求項２記載の原子
炉炉心流量計算方法及び炉心冷却材流量測定装置。3. The reactor core flow rate calculation method according to claim 2, wherein in the core coolant flow rate calculation, an input constant corresponding to the cladding thickness of the fuel surface is adjusted to correct the core coolant flow rate. And core coolant flow rate measuring device.
の摩擦係数を調整し、炉心冷却材流量を補正することを
特徴とする請求項２記載の原子炉炉心流量計算方法及び
炉心冷却材流量測定装置。4. The reactor core flow rate calculation method and the core coolant flow rate measurement according to claim 2, wherein in the core coolant flow rate calculation, the coefficient of friction of the fuel surface is adjusted to correct the core coolant flow rate. apparatus.
面粗さに対応した入力定数を調整し、炉心冷却材流量を
補正することを特徴とする請求項２記載の原子炉炉心流
量計算方法及び炉心冷却材流量測定装置。5. The reactor core flow rate calculation method according to claim 2, wherein in the core coolant flow rate calculation, an input constant corresponding to the surface roughness of the fuel is adjusted to correct the core coolant flow rate. Core coolant flow rate measuring device.
炉周りの熱収支から算出されるヒートバランス流量にフ
ィルター処理を行ない、燃料へのクラッド付着等の変化
時間の長い不確定要因の影響を受けない長期トレンド流
量を算出することを特徴とする請求項２〜５記載の原子
炉炉心流量方法及び炉心冷却材流量測定装置。6. The long-term trend flow rate calculator performs a filtering process on a heat balance flow rate calculated from a heat balance around the reactor, and is not affected by uncertain factors having a long change time such as clad adhesion to fuel. 6. The reactor core flow rate method and the core coolant flow rate measuring apparatus according to claim 2, wherein a long-term trend flow rate is calculated.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP32503399A JP2001141874A (en) | 1999-11-16 | 1999-11-16 | Flow rate-measuring device for reactor core cooling material |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP32503399A JP2001141874A (en) | 1999-11-16 | 1999-11-16 | Flow rate-measuring device for reactor core cooling material |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JP2001141874A true JP2001141874A (en) | 2001-05-25 |
Family
ID=18172410
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP32503399A Pending JP2001141874A (en) | 1999-11-16 | 1999-11-16 | Flow rate-measuring device for reactor core cooling material |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JP2001141874A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2011122910A (en) * | 2009-12-10 | 2011-06-23 | Toshiba Corp | Nuclear reactor and core flow evaluation equipment |
-
1999
- 1999-11-16 JP JP32503399A patent/JP2001141874A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2011122910A (en) * | 2009-12-10 | 2011-06-23 | Toshiba Corp | Nuclear reactor and core flow evaluation equipment |
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