JPH0525313B2 - - Google Patents
Info
- Publication number
- JPH0525313B2 JPH0525313B2 JP19872488A JP19872488A JPH0525313B2 JP H0525313 B2 JPH0525313 B2 JP H0525313B2 JP 19872488 A JP19872488 A JP 19872488A JP 19872488 A JP19872488 A JP 19872488A JP H0525313 B2 JPH0525313 B2 JP H0525313B2
- Authority
- JP
- Japan
- Prior art keywords
- dose equivalent
- neutron
- proportional counter
- energy
- reaction
- 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
- -1 hydrogen compound Chemical class 0.000 claims description 18
- 239000001257 hydrogen Substances 0.000 claims description 9
- 229910052739 hydrogen Inorganic materials 0.000 claims description 9
- 239000004698 Polyethylene Substances 0.000 claims description 8
- 229920000573 polyethylene Polymers 0.000 claims description 8
- WTEOIRVLGSZEPR-UHFFFAOYSA-N boron trifluoride Chemical compound FB(F)F WTEOIRVLGSZEPR-UHFFFAOYSA-N 0.000 claims description 4
- ZOXJGFHDIHLPTG-UHFFFAOYSA-N Boron Chemical compound [B] ZOXJGFHDIHLPTG-UHFFFAOYSA-N 0.000 claims description 3
- 229910052796 boron Inorganic materials 0.000 claims description 3
- 229910015900 BF3 Inorganic materials 0.000 claims description 2
- 239000002245 particle Substances 0.000 description 9
- 230000005855 radiation Effects 0.000 description 5
- 239000007789 gas Substances 0.000 description 4
- 238000010586 diagram Methods 0.000 description 3
- 238000001514 detection method Methods 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 238000005259 measurement Methods 0.000 description 2
- 230000035945 sensitivity Effects 0.000 description 2
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- 231100000987 absorbed dose Toxicity 0.000 description 1
- XAGFODPZIPBFFR-UHFFFAOYSA-N aluminium Chemical compound [Al] XAGFODPZIPBFFR-UHFFFAOYSA-N 0.000 description 1
- 229910052782 aluminium Inorganic materials 0.000 description 1
- 239000004020 conductor Substances 0.000 description 1
- 238000007796 conventional method Methods 0.000 description 1
- 231100000673 dose–response relationship Toxicity 0.000 description 1
- 238000010894 electron beam technology Methods 0.000 description 1
- 150000002483 hydrogen compounds Chemical class 0.000 description 1
- 239000012528 membrane Substances 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 229910001220 stainless steel Inorganic materials 0.000 description 1
- 239000010935 stainless steel Substances 0.000 description 1
Landscapes
- Measurement Of Radiation (AREA)
Description
【発明の詳細な説明】
〔産業上の利用分野〕
本発明は粒子加速器取扱施設、核燃料施設、原
子炉施設等における放射線計測、放射線管理に用
いることのできる中性子検出器に関するものであ
る。DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to a neutron detector that can be used for radiation measurement and radiation management in particle accelerator handling facilities, nuclear fuel facilities, nuclear reactor facilities, and the like.
〔従来の技術〕
一般に、生体に対する放射線の効果を評価する
場合、線量当量という概念が用いられ、吸収線量
と線質係数の積で与えられる。線質係数は、粒子
の飛跡に沿つた生体分子に対する局所的なエネル
ギーの付加率、即ち線エネルギー付与と共に増加
し、高速電子線、ベータ線、Ｘ線、ガンマ線等で
は１であり、α粒子、中性子等では大きな値とな
る。また線量当量は、実効線量当量や組織線量当
量に分類され、線量当量率線量当量の単位時間当
たりの値として定義される。[Prior Art] Generally, when evaluating the effects of radiation on a living body, the concept of dose equivalent is used, and is given as the product of absorbed dose and radiation quality coefficient. The radiation quality coefficient is the local energy addition rate to biomolecules along the particle trajectory, that is, it increases with the addition of linear energy, and is 1 for high-speed electron beams, beta rays, X-rays, gamma rays, etc., and is 1 for α particles, This is a large value for neutrons, etc. Dose equivalents are classified into effective dose equivalents and tissue dose equivalents, and are defined as the value of dose equivalent rate per unit time.
従来、中性子線量当量率の測定は、中性子をポ
リエチレン等の中性子減速材で減速させ、３フツ
カ硼素（BF3）を封入したアルミ、或いはステン
レス等からなる容器中に入射させて（ｎ、α）反
応を起こさせ、このとき出るα粒子（α）でガス
をイオン化して、これを電気信号パルスとして出
力端子より取り出すようにしていた。 Conventionally, the neutron dose equivalent rate has been measured by moderating neutrons with a neutron moderator such as polyethylene, and making them enter a container made of aluminum or stainless steel containing 30% boron (BF 3 ) (n, α). A reaction was caused, and the gas was ionized by the α particles (α) produced during this reaction, which were then extracted from the output terminal as electrical signal pulses.
しかしながら、従来の中性子検出器では線量応
答特性が広いエネルギー領域で一定となるように
ポリエチレン等の中性子減速材を多く使用してい
るため、装置が大型化しかつ重くなりがちで、そ
のため検出器の小型軽量化を図る上で障害になつ
ていた。また、この方式の線量当量率計では実効
線量当量や組織線量当量等の異なつた線量当量を
一つの検出器で測定することが不可能であると共
に、中性子の平均エネルギー等のエネルギー情報
を得ることができなかつた。
However, conventional neutron detectors use a large amount of neutron moderating material such as polyethylene so that the dose response characteristics remain constant over a wide energy range, which tends to make the equipment large and heavy. This has become an obstacle in efforts to reduce weight. In addition, with this type of dose equivalent rate meter, it is impossible to measure different dose equivalents such as effective dose equivalent and tissue dose equivalent with one detector, and it is also difficult to obtain energy information such as the average energy of neutrons. I couldn't do it.
本発明は上記問題点を解決するためのもので、
中性子の実効線量当量や組織線量当量等、エネル
ギーにより応答特性が異なる線量当量を同時に測
定でき、かつ中性子の平均エネルギーも求めるこ
とができ、中性子サーベイメータまたはモニタの
小型軽量化を図ることの可能な中性子検出器を提
供することを目的とする。 The present invention is intended to solve the above problems,
A neutron sensor that can simultaneously measure dose equivalents that have different response characteristics depending on energy, such as the effective dose equivalent and tissue dose equivalent of neutrons, and also determine the average energy of neutrons, making it possible to reduce the size and weight of neutron survey meters or monitors. The purpose is to provide a detector.
そのために本発明の中性子検出器は、３フツ化
ほう素を充填し、ほう素の（ｎ、α）反応を利用
したBF3比例計数管を用いた中性子検出器におい
て、比例計数管の容器の内面にポリエチレン等の
水素化合物の膜をコーテイングしたこと、比例計
数管の出力を波高または波形弁別回路により弁別
し、２種類以上の電気信号を取り出すようにし、
また波高または波形弁別回路により弁別された２
種類以上の電気信号を各々計数し、各計数値の比
から線量当量、平均エネルギーを求めるようにし
たことを特徴とする。
To this end, the neutron detector of the present invention uses a BF 3 proportional counter filled with boron trifluoride and utilizes the (n, α) reaction of boron. The inner surface is coated with a film of hydrogen compound such as polyethylene, the output of the proportional counter is discriminated by a wave height or waveform discrimination circuit, and two or more types of electrical signals are extracted.
Also, the wave height or waveform discrimination circuit discriminates 2
It is characterized in that more than one type of electrical signal is counted, and the dose equivalent and average energy are determined from the ratio of each counted value.
本発明の中性子検出器はBF3比例計数管の容器
の内面にポリエチレン等の水素化合物の膜をコー
テイングし、低速中性子においては（ｎ、α）反
応により、高速中性子においては（ｎ、ｐ）反応
によりそれぞれ電気信号を取り出し、（ｎ、α）
反応、（ｎ、ｐ）反応により波高または波形が異
なることを利用し、それぞれ分離して計数し、そ
の計数比と線量当量、平均エネルギーとの間に所
定の函数関係があることを利用して線量当量や平
均エネルギー等も求めることができるため、１つ
の検出器で実効線量当量や組織線量当量の測定を
同時に行うことができ、またポリエチレン等の水
素化合物が高速中性子に反応するため多量の減速
材を必要とせず、装置の小型軽量化を図ることが
できる。
The neutron detector of the present invention coats the inner surface of the container of the BF 3 proportional counter with a film of a hydrogen compound such as polyethylene. Extract the electrical signals respectively by (n, α)
Taking advantage of the fact that the wave height or waveform differs depending on the reaction or (n,p) reaction, each is counted separately, and the fact that there is a predetermined functional relationship between the counting ratio, dose equivalent, and average energy. Since dose equivalent and average energy can also be determined, effective dose equivalent and tissue dose equivalent can be measured simultaneously with one detector. Also, since hydrogen compounds such as polyethylene react with fast neutrons, large amounts of deceleration occur. No materials are required, and the device can be made smaller and lighter.
〔実施例〕 以下、実施例を図面を参照して説明する。〔Example〕 Examples will be described below with reference to the drawings.
第１図は本発明の中性子検出器の構造を示す図
で、図中、１０は比例計数管、１１は計数管容
器、１２は中心電極用芯線、１３は出力端子、１
４は水素化合物の膜、１５は外側電極である。 FIG. 1 is a diagram showing the structure of the neutron detector of the present invention, in which 10 is a proportional counter tube, 11 is a counter tube container, 12 is a core wire for the center electrode, 13 is an output terminal, 1
4 is a hydrogen compound film, and 15 is an outer electrode.
図において、比例計数管１０は容器１１の内面
にポリエチレン等の水素化合物の膜１４がコーテ
イングされ、その膜の表面に導電体を塗布して外
側電極１５を形成し、中心電極用芯線１２との間
に高電圧が印加されており、内部にはBF3ガスが
封入されている。 In the figure, a proportional counter 10 has a container 11 whose inner surface is coated with a film 14 of a hydrogen compound such as polyethylene, a conductor applied to the surface of the film to form an outer electrode 15, and a core wire 12 for a center electrode. A high voltage is applied between them, and BF 3 gas is sealed inside.
そしてエネルギーの低い低速中性子が入射する
と（ｎ、α）反応が起こり、比例計数管内にα粒
子が放出される。このα粒子によりガスがイオン
化され、イオン化粒子が円筒形または球形の外側
電極１５と、中心電極である芯線１２の間の高電
圧により加速されて両電極間に電流が流れて増幅
され、出力端子１３より電気信号パルスとして出
力される。また、高速中性子が入射した場合に
は、比例計数管１０の容器１１の内側にコーテイ
ングしたポリエチレン等の水素化合物の膜の１４
中の水素と、（ｎ、ｐ）反応を起こし、この反跳
陽子(p)が比例計数管内に放出される。この反跳陽
子がα粒子の場合と同様にガスをイオン化させ、
電気信号パルスとして出力端子より取り出され
る。そして、低速中性子と高速中性子が入射した
場合の出力信号は電圧レベルや波形で異なるた
め、信号弁別回路で弁別可能で、それぞれ別々に
計数することができる。 When low-energy slow neutrons are incident, an (n, α) reaction occurs, and α particles are released into the proportional counter. The gas is ionized by these α particles, and the ionized particles are accelerated by the high voltage between the cylindrical or spherical outer electrode 15 and the core wire 12, which is the center electrode, and a current flows between the two electrodes and is amplified. 13, it is output as an electrical signal pulse. In addition, when fast neutrons are incident, the hydrogen compound film coated on the inside of the container 11 of the proportional counter 10
A (n,p) reaction occurs with the hydrogen inside, and the recoil protons (p) are released into the proportional counter tube. These recoil protons ionize the gas in the same way as alpha particles,
It is extracted from the output terminal as an electrical signal pulse. Since the output signals when slow neutrons and fast neutrons are incident differ in voltage level and waveform, they can be discriminated by a signal discrimination circuit and each can be counted separately.
第２図はこのような弁別回路を有する本発明の
検出回路構成を示す図で、図中、２１は高電圧回
路、２２は増幅器、２３は信号弁別回路、２４，
２５は計数回路、２６は演算処理回路、２７は表
示部である。 FIG. 2 is a diagram showing the configuration of a detection circuit of the present invention having such a discrimination circuit, in which 21 is a high voltage circuit, 22 is an amplifier, 23 is a signal discrimination circuit, 24,
25 is a counting circuit, 26 is an arithmetic processing circuit, and 27 is a display section.
比例計数管１０には高電圧回路２１により外側
電極と中心電極間に高電圧が印加されており、前
述したように中性子が入射すると出力端子より電
気信号パルスが取り出され、この出力パルスを増
幅器２２で増幅し、信号弁別回路２３でエネルギ
ーレベル或いは波形によつて、α信号（α粒子に
より生じた信号）、或いはｐ信号（反跳陽子によ
り生じた信号）に弁別すし、それぞれ計数回路２
４，２５により計数する。 A high voltage is applied to the proportional counter tube 10 between the outer electrode and the center electrode by the high voltage circuit 21, and as described above, when a neutron is incident, an electric signal pulse is extracted from the output terminal, and this output pulse is sent to the amplifier 22. The signal is amplified by a signal discriminator circuit 23 and discriminated into an α signal (signal generated by α particles) or a p signal (signal generated by recoil protons) depending on the energy level or waveform, and a counting circuit 2
Count by 4,25.
ところで、中性子エネルギーに対する（ｎ、
ｐ）反応、（ｎ、α）反応の感度は第３図のよう
な特性を示し、各反応は中性子のエネルギーによ
りそれぞれ異なり、所定の確率で生じている。そ
こで、第２図の計数回路２４における計数値を
C2、計数回路２５における計数値をC1とし、単
一エネルギー中性子による照射実験や中性子輸送
計算による感度解析を行うことにより、計数比
C2／C1と線量当量Ｄ、又は平均エネルギーEave
との間に、例えば第４図に示すような函数曲線が
求められる。なお、図の函数曲線は計数値C1で
除して規格化している。この函数をプログラム化
し、演算処理回路２６に記憶させておき、任意の
中性子エネルギー分布を持つ場所で測定すること
により、計数値C1およびC2から第４図の函数関
係を利用し、演算処理回路２６により線量当量や
平均エネルギーを算出することができる。 By the way, (n,
The sensitivities of the p) reaction and the (n, α) reaction exhibit characteristics as shown in FIG. 3, and each reaction differs depending on the energy of the neutron and occurs with a predetermined probability. Therefore, the count value in the counting circuit 24 in FIG.
C2, the count value in the counting circuit 25 is set as C1, and the count ratio is
C2/C1 and dose equivalent D, or average energy E ave
For example, a functional curve as shown in FIG. 4 is obtained between . Note that the function curve in the figure is normalized by dividing by the count value C1. By programming this function, storing it in the arithmetic processing circuit 26, and measuring it at a location with an arbitrary neutron energy distribution, the arithmetic processing circuit 26 can utilize the functional relationship shown in FIG. The dose equivalent and average energy can be calculated by
こうして演算処理回路２２で線量当量や平均エ
ネルギーが算出され、その値が表示部２７に表示
される。 In this way, the arithmetic processing circuit 22 calculates the dose equivalent and the average energy, and the values are displayed on the display section 27.
以上のように本発明によれば、１つの検出器で
実効線量当量や組織線量当量の測定を同時に行う
ことが可能となり、また測定場所の平均エネルギ
ー等のエネルギー情報を得ることが可能となる。
そしてポリエチレン等の水素化合物が高速中性子
に反応するため、従来のように多量の減速材を必
要としないため、中性子検出器の小型軽量化を図
ることができる。
As described above, according to the present invention, it is possible to simultaneously measure the effective dose equivalent and the tissue dose equivalent with one detector, and it is also possible to obtain energy information such as the average energy of the measurement location.
Since a hydrogen compound such as polyethylene reacts with fast neutrons, a large amount of moderator is not required as in the conventional method, so the neutron detector can be made smaller and lighter.
第１図は本発明の中性子検出器の構造を示す
図、第２図は本発明の中性子検出回路の構成を示
す図、第３図は中性子エネルギーに対する反応の
感度特性を示す図、第４図は計数比C2／C1と線
量当量Ｄ、または平均エネルギーEaveとの関係を
示す図である。
１０……比例計数管、１１……計数管容器、１
２……中心電極用芯線、１３……出力端子、１４
……水素化合物の膜、１５……外側電極、２１…
…高電圧回路、２２……増幅器、２３……信号弁
別回路、２４，２５……計数回路、２６……演算
処理回路、２７……表示部。
Fig. 1 shows the structure of the neutron detector of the present invention, Fig. 2 shows the structure of the neutron detection circuit of the invention, Fig. 3 shows the sensitivity characteristics of the reaction to neutron energy, and Fig. 4 is a diagram showing the relationship between the counting ratio C2/C1 and the dose equivalent D or the average energy E ave . 10...Proportional counter tube, 11...Counter tube container, 1
2...Core wire for center electrode, 13...Output terminal, 14
... Hydrogen compound membrane, 15 ... Outer electrode, 21 ...
...High voltage circuit, 22...Amplifier, 23...Signal discrimination circuit, 24, 25...Counting circuit, 26...Arithmetic processing circuit, 27...Display section.
Claims (1)
α）反応を利用したBF3比例計数管を用いた中性
子検出器において、比例計数管の容器の内面にポ
リエチレン等の水素化合物の膜をコーテイングし
たことを特徴とする中性子検出器。 ２ 比例計数管の出力を波高または波形弁別回路
により弁別し、２種類以上の電気信号を取り出す
ようにした請求項１記載の中性子検出器。 ３ 波高または波形弁別回路により弁別された２
種類以上の電気信号を各々計数し、各計数値の比
から線量当量、平均エネルギーを求めるようにし
た請求項２記載の中性子検出器。[Claims] 1 Filled with boron trifluoride, boron (n,
α) A neutron detector using a BF 3 proportional counter that utilizes a reaction, characterized in that the inner surface of the container of the proportional counter is coated with a film of a hydrogen compound such as polyethylene. 2. The neutron detector according to claim 1, wherein the output of the proportional counter is discriminated by a wave height or waveform discrimination circuit to extract two or more types of electrical signals. 3 Discriminated by wave height or waveform discrimination circuit 2
3. The neutron detector according to claim 2, wherein more than one type of electrical signal is counted, and the dose equivalent and average energy are determined from the ratio of each counted value.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP19872488A JPH0247581A (en) | 1988-08-09 | 1988-08-09 | Neutron detector |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP19872488A JPH0247581A (en) | 1988-08-09 | 1988-08-09 | Neutron detector |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH0247581A JPH0247581A (en) | 1990-02-16 |
| JPH0525313B2 true JPH0525313B2 (en) | 1993-04-12 |
Family
ID=16395938
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP19872488A Granted JPH0247581A (en) | 1988-08-09 | 1988-08-09 | Neutron detector |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0247581A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2001255378A (en) * | 2000-03-13 | 2001-09-21 | Natl Inst For Fusion Science | Radiation detector |
Families Citing this family (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP5126739B2 (en) * | 2007-10-16 | 2013-01-23 | 大学共同利用機関法人 高エネルギー加速器研究機構 | Gas detector for neutron measurement |
| JP2010223632A (en) * | 2009-03-19 | 2010-10-07 | National Institute Of Advanced Industrial Science & Technology | Neutron energy measuring instrument |
| JP5868256B2 (en) * | 2012-04-26 | 2016-02-24 | 三菱電機株式会社 | Dose rate measuring device |
-
1988
- 1988-08-09 JP JP19872488A patent/JPH0247581A/en active Granted
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2001255378A (en) * | 2000-03-13 | 2001-09-21 | Natl Inst For Fusion Science | Radiation detector |
Also Published As
| Publication number | Publication date |
|---|---|
| JPH0247581A (en) | 1990-02-16 |
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