JPS62174693A - Method of measuring output from nuclear reactor - Google Patents
Method of measuring output from nuclear reactorInfo
- Publication number
- JPS62174693A JPS62174693A JP61015883A JP1588386A JPS62174693A JP S62174693 A JPS62174693 A JP S62174693A JP 61015883 A JP61015883 A JP 61015883A JP 1588386 A JP1588386 A JP 1588386A JP S62174693 A JPS62174693 A JP S62174693A
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- Prior art keywords
- reactor
- core
- power distribution
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- detectors
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Links
- 238000000034 method Methods 0.000 title description 16
- 238000009826 distribution Methods 0.000 claims description 31
- 238000000691 measurement method Methods 0.000 claims description 6
- 239000011159 matrix material Substances 0.000 description 13
- 230000009466 transformation Effects 0.000 description 8
- 238000004364 calculation method Methods 0.000 description 5
- 238000004458 analytical method Methods 0.000 description 3
- 238000006243 chemical reaction Methods 0.000 description 2
- 238000007796 conventional method Methods 0.000 description 2
- 230000007423 decrease Effects 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 238000004519 manufacturing process Methods 0.000 description 2
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 1
- 238000003780 insertion Methods 0.000 description 1
- 230000037431 insertion Effects 0.000 description 1
- 238000005259 measurement Methods 0.000 description 1
Classifications
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T3/00—Measuring neutron radiation
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
- G21C17/10—Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
-
- 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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- Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- Health & Medical Sciences (AREA)
- Life Sciences & Earth Sciences (AREA)
- General Physics & Mathematics (AREA)
- Molecular Biology (AREA)
- Spectroscopy & Molecular Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
〔産業上の利用分野〕
この発明は、原子炉出力測定方法ｌと関し、さらｌこ詳
しくいうと、中性子検出器を用いて、原子炉炉心内の出
力分布を測定するための原子炉出力測定方法に関するも
のである。[Detailed Description of the Invention] [Industrial Application Field] The present invention relates to a method for measuring nuclear reactor power, and more specifically, a method for measuring power distribution in a nuclear reactor core using a neutron detector. This article relates to a method for measuring nuclear reactor output.
Ｃ従来の技術〕
原子炉の炉心内出方分布は原子炉を運転する上で重要な
情報であり、従来からこれを測定するいくつかの方法が
試みられている。その一つは原子炉内に多数の定置形の
小形中性子検出器を設置する方法である。この方法では
、検出器の故障時の交換が困難なので、長期間安定に動
作する検出器を必要とするが、そのような検出器を製作
することはきわめて困難であった。第２の方法として、
小数の小形中性子検出器を、原子炉内に設けた多数の検
出器挿入孔（こ順次に挿入し、走査しながら炉出力分布
を測定する方法がある。この方法では原子炉全体の炉出
力分布を測定するのに時間ががかり、炉出力分布が短時
間で変化する場合に追従で音ないという欠点があった。C. Prior Art] The distribution in the core of a nuclear reactor is important information for operating a nuclear reactor, and several methods have been tried to measure this. One method is to install a large number of stationary small neutron detectors inside the reactor. This method requires a detector that operates stably for a long period of time since it is difficult to replace the detector when it breaks down, but it has been extremely difficult to manufacture such a detector. As a second method,
There is a method in which a small number of small neutron detectors are inserted into a number of detector insertion holes (sequentially) in the reactor and the reactor power distribution is measured while scanning. It takes a long time to measure, and there is a drawback that there is no sound when the furnace power distribution changes in a short period of time.
第３の方法として、加圧水膨軽水炉（ＰＷＲ）の炉出力
測定用として従来から使廟されている炉外設置形の中性
子計測装置の検出器を軸方向に分割して多数の短尺検出
器とし、各検出器からの出力信号を用いて計算によって
炉出力分布を求める方法がある。As a third method, the detector of an external neutron measuring device, which has been used conventionally for measuring the reactor power of a pressurized water expansion light water reactor (PWR), is divided in the axial direction into a number of short detectors. There is a method of calculating the reactor power distribution using output signals from each detector.
上記の第３の方法について、以下に第２図を用いて説明
する。図において、（１００）はＰＷＲ形原子炉の炉心
、（１０１）〜（１０４）は軸方向に分割した原子炉炉
心の各部分、（２１０）（２２０）（２３０）（２４０
）はそれぞれ中性子検出器、（２４１）〜（２１４）
、 （２２１）〜（２１４）。The third method described above will be explained below using FIG. 2. In the figure, (100) is the core of a PWR type reactor, (101) to (104) are each part of the reactor core divided in the axial direction, (210) (220) (230) (240)
) are neutron detectors, (241) to (214) respectively.
, (221)-(214).
（２３１）〜（２３４）および（２４１）〜（２４４）
はそれぞれ中性子検出器（２１０）（２２０）（２５０
）および（２４０）の分割された部分検出器を示す。(231) to (234) and (241) to (244)
are neutron detectors (210), (220), and (250), respectively.
) and (240) are shown.
原子炉炉心内には、図示していないが、前記の第２の方
法として説明した走査形の炉心内中性子検出器が設けら
れており、原子炉の炉心内出力分布ｆ（ｘ、ｙ、ｚ）が
求められる。第２図に示す装置は、この炉出力分布ｆ
（Ｘ、７．Ｚ）の！７平面（水平面）に関する平均値ｆ
（ｚ）（上下方向の出力分布）を求めるものである。Although not shown in the reactor core, the scanning type in-core neutron detector explained as the second method is installed, and the in-core power distribution f(x, y, z ) is required. The device shown in Fig. 2 has this furnace power distribution f
(X, 7.Z)! Average value f for 7 planes (horizontal plane)
(z) (output distribution in the vertical direction) is determined.
（Ｓｚア：炉心の水平方向断面積）
ここで、ｆ（ｚ）はフーリエ級数の和として次のように
表わすことができる。(Sza: Horizontal cross-sectional area of core) Here, f(z) can be expressed as a sum of Fourier series as follows.
（２菖：炉心の高さ）
ただし、炉出力分布の一般的な形状が炉心上下端で減少
する形となるので、０〜πを変域とするサイン系列で表
わしている。係数０１は軸方向出力分布ｆ（ｚ）から求
められるが、ｆ （ｚ）からＣ１を求める関係式はいわ
ゆるフーリエ級数展開であり、次式で表わされる。(2 iris: height of the core) However, since the general shape of the reactor power distribution is such that it decreases at the upper and lower ends of the core, it is expressed as a sine series with a range of 0 to π. The coefficient 01 is obtained from the axial output distribution f(z), and the relational expression for obtaining C1 from f(z) is a so-called Fourier series expansion, and is expressed by the following equation.
ここで、この装置では各部分検出器（２１１”）〜（２
１４）等の出力信号値から上記の係数を決定することが
必要となるので、式（３）の代りに次式を用いると都合
がよい。Here, in this device, each partial detector (211'') to (2
Since it is necessary to determine the above coefficients from the output signal values such as 14), it is convenient to use the following equation instead of equation (3).
（Ｃｔ） ＝ （Ａｔｊ）（ｆ、１） ・
・・・・・・（４）ここに
〔Ａｉｊ〕は定係数行列であり、式（２）を区間１毎に
積分した結果を解いて得られる。ｆｊはｆ （ｚ）を原
子炉部分（１０１）〜（１０４）の各々につきｚ軸方向
に積分したものであるが、部分検出器（２１１）〜（２
１４）等の出力とは次式の関係にある。(Ct) = (Atj) (f, 1) ・
(4) Here, [Aij] is a constant coefficient matrix, which is obtained by solving the result of integrating equation (2) for each interval. fj is the integral of f (z) in the z-axis direction for each of the reactor sections (101) to (104), but
14) etc., there is a relationship as shown in the following equation.
（Ｄｋ） ＝ （Ｑ＋ｃ５ ） Ｃｆ、１ ）
・・・・・・・（５）ここでＤｋは、部分検出器（
２１１）（２２１）（２３１）（２４１）の平均につい
てＤｌ、（２１？）（２２２）（２３２）（２４２）の
平均についてＤ２ Ｃ以下同様）のように定義する。(Dk) = (Q+c5) Cf, 1)
(5) Here, Dk is the partial detector (
Define Dl for the average of 211)(221)(231)(241), and D2 for the average of (21?)(222)(232)(242).
（Ｑｋ４ ）は炉心各部の出力が各部分検出器に与える
寄与の比率を係数とした定係数行列である。(Qk4) is a constant coefficient matrix whose coefficient is the ratio of the contribution of the output of each part of the core to each partial detector.
式（４）と式（５）からフーリエ係数０１が次のように
求められる。Fourier coefficient 01 is obtained from equations (4) and (5) as follows.
ＣＯ１〕＝ （Ａｉｊ）（Ｑｋ５） （Ｄｃ）
・・・・・（６）さらに式（２）によって炉心軸方向出
力分布ｆ （ｚ）が求められる。CO1] = (Aij) (Qk5) (Dc)
...(6) Furthermore, the core axial power distribution f (z) is determined by equation (2).
以上のような従来の原子炉出力測定方法では、式（５）
の定係数行列（’Ｌ＋ｃｊ、１を決定するためには炉出
力分布の異った状態に対して検出器出力〔Ｄｋ〕と原子
炉部分出力積分値ｆｊの組を多数用意する必要がある。In the conventional reactor power measurement method as described above, Equation (5)
In order to determine the constant coefficient matrix ('L+cj, 1), it is necessary to prepare many pairs of detector output [Dk] and reactor partial power integral value fj for different states of the reactor power distribution.
この組の数は、少なくとも部分検出器の分割数（第２図
では４）以上でなければならない。また、各組の炉出力
分布は十分に異ったものでなければ、行列（’Ｑｃｊ）
の算出が困難となるなどの問題点があった。The number of these sets must be at least equal to or greater than the number of divisions of the partial detector (4 in FIG. 2). In addition, unless the reactor power distributions of each set are sufficiently different, the matrix ('Qcj)
There were problems such as difficulty in calculating .
この発明はかような問題点を解消しようとするもので、
変換行列（Ｑｋｊ）の決定を容易にし、炉出力分布演算
の精度を向上し、かつ、炉出力分布情報の詳細化を行う
ことができる原子炉出力測定方法を得ることを目的とす
る。This invention attempts to solve these problems,
It is an object of the present invention to obtain a reactor power measurement method that facilitates the determination of a transformation matrix (Qkj), improves the accuracy of reactor power distribution calculation, and allows detailed reactor power distribution information.
この発明に係る原子炉出力測定方法は、変換行列（Ｑｋ
ｊ）を求める際にｆｊと〔Ｄｋ〕との対応を求めること
をせず、ｆｊの代りに各中性子検出器それぞれの正面に
位置する方位角の区画内の炉心内情報のみを変換行列の
作成に利用する。The reactor power measurement method according to the present invention includes a transformation matrix (Qk
When calculating j), instead of calculating the correspondence between fj and [Dk], instead of fj, only the in-core information in the azimuth angle section located in front of each neutron detector is used to create a conversion matrix. Use it for.
この発明においては、変換行列（Ｑｃｊ）を求めるため
のデータの組が有する相互相関が高くなり、（Ｑ＋ｃｊ
）の算出が容易となる。また、中性子検出器それぞれに
対向する方位角領域毎の軸方向炉出力分布が求められる
ので、炉出力分布情報が詳細化できる。In this invention, the cross-correlation of the data set for determining the transformation matrix (Qcj) becomes high, and (Q+cj
) becomes easy to calculate. Furthermore, since the axial reactor power distribution is determined for each azimuth region facing each neutron detector, the reactor power distribution information can be detailed.
以下、この発明の一実施例を第１図を参照して説明する
。ここで、第２図に示す従来技術によるものと異る点は
、原子炉炉心を方位角方向に分割し、それぞれ（１１１
）〜（１１４） 、 （１２１）〜（１２４）、・・・
・・・のように１６個の炉心部分に分割して解析を行う
点である。解析手順の違いは、式（５）において、変換
行列（Ｑｋ６）を求める場合に、原子炉部分出力積分値
で、の代りに各検出器の正面に対向する１／４方位角部
分に関する原子炉部分出力積分値を用いる点である。す
なわち、４つの検出器（２１０） （２２０）（２３０
）（２４０）をそれぞれり、〜Ｄ４で表わし、それぞれ
の部分検出器出力をＤ＋ｋ””４にで表わし、炉心部分
（１１１’）〜（１１４）の部分出力積分値をｆ７５、
炉心部分（１２１）〜（１２４）につきｆ２．のようｌ
こ表わし、式（５）を下記の４つの式に分解する。An embodiment of the present invention will be described below with reference to FIG. Here, the difference from the prior art shown in FIG. 2 is that the reactor core is divided in the azimuth direction, and each (111
)~(114), (121)~(124),...
The point is that the analysis is performed by dividing the reactor core into 16 core parts. The difference in the analysis procedure is that in formula (5), when calculating the transformation matrix (Qk6), the reactor partial power integral value is used instead of the reactor partial power integral value for the 1/4 azimuth angle portion facing the front of each detector. The point is to use partial output integral values. That is, four detectors (210) (220) (230
) (240) are respectively represented by ~D4, each partial detector output is represented by D+k""4, and the partial power integral value of the core portions (111') to (114) is represented by f75,
f2. for core portions (121) to (124). Like l
Therefore, equation (5) is decomposed into the following four equations.
ＣＤｔｋ〕＝（Ｑ＋ｋｊ） （ｆ＋ｊ’） ・・・
・・（５−１）（Ｄｚｋ） ＝ （Ｑｚｋｊ） （ｆ２
ｊ ） ・・・・・（ｓ−２）（Ｄｓ＋ｃ） ＝
（Ｑｓｋ、１）Ｃｆａｊ） ・−・（５−３）〔Ｄ
４ｋ）″（Ｑ４ｋｊ） Ｃｆ４ｊ） …・・（５−
４）式（Ｓ−１）〜（５−４）のそれぞれにつき４つの
変換行列（’Ｌｒ５Ｂ）〜（Ｑ４ｋｊ）を求め、それぞ
れ式（６）に代入して得られるＣ１を０１１〜ｃ４ｉと
する。Ｃｊｉ〜Ｃ４１のそれぞれを式（２）に代入して
得られる炉出力分布は、各検出器（２１０）〜（２４０
）の正面に対向する１／４方位角炉心部分それぞれの軸
方向炉出力分布である。これ等をｆ、（ｚ）〜ｆ４（ｚ
）とすればｆ（ｚ）は次式で求められ、形態的には従来
技術と同様の最終結果ｆ （Ｚ）が得られる。CDtk]=(Q+kj) (f+j')...
...(5-1)(Dzk) = (Qzkj) (f2
j ) ...(s-2)(Ds+c) =
(Qsk, 1) Cfaj) ・-・(5-3) [D
4k)''(Q4kj) Cf4j)...(5-
4) Find four transformation matrices ('Lr5B) to (Q4kj) for each of formulas (S-1) to (5-4), and substitute them into formula (6) to obtain C1 as 011 to c4i. . The reactor power distribution obtained by substituting each of Cji to C41 into equation (2) is given by each detector (210) to (240).
) is the axial reactor power distribution for each of the 1/4 azimuth angle core sections facing the front. These are expressed as f, (z) ~ f4 (z
), then f(z) is obtained by the following equation, and the final result f(Z) is obtained which is morphologically similar to that of the prior art.
次に前記のように解析手順を変更したことによる利点に
ついて説明する。従来技術において、炉出力分布の演算
精度を決定するものは式（５）の変換行列（Ｑｃｊ）の
数値精度であり、限られた測定ケースにもとづく〔Ｄｋ
〕と〔ｆ、〕の組からＣＱ＋ｇ）を高精度で求めること
１こ困難の原因があった。ここで原子炉炉心の各部分の
出力と特定の部分検出器の出力信号値との関係を考える
と、水平面方向には検出器の近傍の炉心部分の出力との
相関が高く、水平方向に離れるに従って相関は急激に低
下する。Next, the advantages of changing the analysis procedure as described above will be explained. In the conventional technology, what determines the calculation accuracy of the reactor power distribution is the numerical accuracy of the conversion matrix (Qcj) in equation (5), and it is based on limited measurement cases [Dk
There is one reason why it is difficult to obtain CQ+g) from the set of ] and [f, ] with high precision. Considering the relationship between the output of each part of the reactor core and the output signal value of a specific partial detector, in the horizontal direction there is a high correlation with the output of the core part near the detector, and in the horizontal direction it is far away. The correlation decreases rapidly.
特に熱中性子炉にあっては中性子の平均自由行程が短い
ので、この傾向が著るしい。したがって、式（５）のよ
うに検出器出力および炉心出力ともに水平面内の平均値
同志を比べると距離の遠い炉心部分からの寄与を含んだ
ものを比べることになり、相関が低い。これに対してこ
の発明による計算手順によれば、式（５）の代りに式（
ｓ−＋）〜（５−４）であられされる検出器に接近した
炉心部分の出力のみを取扱うため、各式において〔Ｄ１
ｋ〕と〔ｆｇ）。This tendency is particularly noticeable in thermal neutron reactors because the mean free path of neutrons is short. Therefore, when comparing the average values of both the detector output and the core output in the horizontal plane as in equation (5), the comparison includes contributions from the core portions that are far away, and the correlation is low. On the other hand, according to the calculation procedure according to the present invention, formula (5) is replaced by formula (
In each equation, [D1
k] and [fg).
（１＝１〜４）との相関が高く、変換行列（Ｑｚｋｊ）
を求める計算が容易となり、精度が向上する。また、式
（７）により従来の方法と同一の結果を得られるだけで
なく、４つの方位角のそれぞれについての軸方向炉出力
分布ｆ２（ｚ）（Ｊ、＝１〜４）が中間結果として得ら
れ、従来の方法よりも詳細な炉出力分布が得られる。(1=1 to 4), and the transformation matrix (Qzkj)
It becomes easier to calculate and improve accuracy. Moreover, not only can the same results as the conventional method be obtained using equation (7), but also the axial furnace power distribution f2(z) (J, = 1 to 4) for each of the four azimuth angles can be obtained as an intermediate result. This results in a more detailed furnace power distribution than conventional methods.
なお、上記実施例の説明においては、炉出力分布を表現
するのに式（２）のフーリエ級数による表現を用いたが
、展開に用い得る級数系列はフーリエ級数に限らない。In addition, in the description of the above embodiment, expression using the Fourier series of equation (2) was used to express the reactor power distribution, but the series series that can be used for expansion is not limited to the Fourier series.
エルミート多項式等の他の直交函数系列も有効に利用で
きる。Other orthogonal function sequences such as Hermitian polynomials can also be effectively used.
この発明は、以上の説明から明らかなように、炉出力分
布演算に用いる変換行列を求める場合に、炉心水平面内
平均値の代りに、検出器に接近した分割方位角部分のみ
の平均値を用いたので、検出器出力と炉出力分布との相
関が改善され、変換行列の算出が容易となり、ｃｌｉｌ
定精度が向上する。また、中間結果として各方位角毎の
軸方向炉出力分布が得られるので、炉出力分布がより詳
細にわかるという効果もある。As is clear from the above description, this invention uses the average value of only the divided azimuth angle portion close to the detector, instead of the average value in the core horizontal plane, when calculating the transformation matrix used for reactor power distribution calculation. As a result, the correlation between the detector output and the reactor output distribution has been improved, and the calculation of the transformation matrix has become easier.
Improved accuracy. Furthermore, since the axial furnace power distribution for each azimuth angle is obtained as an intermediate result, there is also the effect that the furnace power distribution can be understood in more detail.
第１図はこの発明の一実施例を説明するための検出器お
よび原子炉炉心の配置を示す斜視図、第２図は従来の原
子炉出力測定方法を説明するための検出器および原子炉
炉心の配置を示す斜視図である。
（１１１）〜（１１４） 、・・・・・（１４１）〜（
１４４）・Φ原子炉炉心（１００）を分割した各炉心部
分、（２１０） （２２０）（２３０）（２４０）・・
中性子検出器、（２１１）〜（２１４） 。
・・・・・（２４１）〜（２４４）・・部分検出器。
なお、各図中、同一符号は同−又は相当部分を示す。FIG. 1 is a perspective view showing the arrangement of a detector and a reactor core for explaining an embodiment of the present invention, and FIG. 2 is a perspective view of a detector and a reactor core for explaining a conventional reactor power measurement method. It is a perspective view showing arrangement of. (111) ~ (114) , ... (141) ~ (
144)・ΦReactor core parts into which the reactor core (100) is divided, (210) (220) (230) (240)...
Neutron detector, (211)-(214). ...(241) to (244)...partial detectors. In each figure, the same reference numerals indicate the same or corresponding parts.
Claims (1)
検出器を配置し、これらの中性子検出器は前記原子炉の
炉心の高さにほぼ等しい長さを有し、かつ、内部に互い
に等しい長さに分割された複数個の部分検出器をそれぞ
れ内蔵し、これらの部分検出器の出力信号を用いて前記
炉心内の鉛直軸方向の出力分布を算出する原子炉出力測
定方法において、 前記炉心を前記中性子検出器に対向する方位角毎に扇形
柱体状に分割し、各前記中性子検出器に含まれる前記部
分検出器の出力信号はそれぞれの前記中性子検出器に対
向する炉心方位角部分の鉛直軸方向炉出力分布の算出に
使用され、前記各方位角部分の鉛直軸方向分布を前記方
位角に関して平均し、前記原子炉の鉛直軸方向炉出力分
布を算出することを特徴とする原子炉出力測定方法。[Scope of Claims] A plurality of neutron detectors are arranged at equal azimuth angles in the vertical direction around the reactor, and these neutron detectors have a length approximately equal to the height of the core of the reactor. and each has a built-in plurality of partial detectors divided into equal lengths, and uses the output signals of these partial detectors to calculate the power distribution in the vertical axis direction within the reactor core. In the reactor power measurement method, the reactor core is divided into fan-shaped columns for each azimuth facing the neutron detector, and the output signal of the partial detector included in each of the neutron detectors is transmitted to each of the neutron detectors. is used to calculate the vertical axial reactor power distribution of the core azimuth angle portion facing the reactor, and the vertical axial distribution of the respective azimuth angle portions is averaged with respect to the azimuth angle to calculate the vertical axial reactor power distribution of the reactor. A nuclear reactor power measurement method characterized by:
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61015883A JPS62174693A (en) | 1986-01-29 | 1986-01-29 | Method of measuring output from nuclear reactor |
| DE19873701457 DE3701457A1 (en) | 1986-01-29 | 1987-01-20 | METHOD FOR MEASURING REACTOR PERFORMANCE |
| FR878700997A FR2597252B1 (en) | 1986-01-29 | 1987-01-28 | METHOD FOR MEASURING THE POWER OF A REACTOR |
| US07/821,871 US5185121A (en) | 1986-01-29 | 1992-01-16 | Reactor power measuring method |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61015883A JPS62174693A (en) | 1986-01-29 | 1986-01-29 | Method of measuring output from nuclear reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS62174693A true JPS62174693A (en) | 1987-07-31 |
| JPH0458916B2 JPH0458916B2 (en) | 1992-09-18 |
Family
ID=11901188
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61015883A Granted JPS62174693A (en) | 1986-01-29 | 1986-01-29 | Method of measuring output from nuclear reactor |
Country Status (3)
| Country | Link |
|---|---|
| JP (1) | JPS62174693A (en) |
| DE (1) | DE3701457A1 (en) |
| FR (1) | FR2597252B1 (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2008175692A (en) * | 2007-01-18 | 2008-07-31 | Nuclear Fuel Ind Ltd | Measuring method of axial power distribution of core |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH07119827B2 (en) * | 1987-12-25 | 1995-12-20 | 三菱電機株式会社 | Measuring device of reactor power distribution |
Family Cites Families (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| DE2346725A1 (en) * | 1973-09-17 | 1975-03-27 | Siemens Ag | NUCLEAR REACTOR |
| DE2515712A1 (en) * | 1975-04-10 | 1976-10-21 | Kraftwerk Union Ag | NUCLEAR REACTOR |
| US4079236A (en) * | 1976-03-05 | 1978-03-14 | Westinghouse Electric Corporation | Method and apparatus for monitoring the axial power distribution within the core of a nuclear reactor, exterior of the reactor |
| FR2546330B1 (en) * | 1983-05-19 | 1985-08-23 | Framatome Sa | METHOD FOR DETECTING FAULTS IN THE CORE POWER DISTRIBUTION OF A PRESSURE WATER NUCLEAR REACTOR AND DEVICE FOR CARRYING OUT SAID METHOD |
-
1986
- 1986-01-29 JP JP61015883A patent/JPS62174693A/en active Granted
-
1987
- 1987-01-20 DE DE19873701457 patent/DE3701457A1/en not_active Ceased
- 1987-01-28 FR FR878700997A patent/FR2597252B1/en not_active Expired - Fee Related
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2008175692A (en) * | 2007-01-18 | 2008-07-31 | Nuclear Fuel Ind Ltd | Measuring method of axial power distribution of core |
Also Published As
| Publication number | Publication date |
|---|---|
| DE3701457A1 (en) | 1987-07-30 |
| FR2597252A1 (en) | 1987-10-16 |
| FR2597252B1 (en) | 1992-01-10 |
| JPH0458916B2 (en) | 1992-09-18 |
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