JP2010276561A - Neutron dosimeter - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a compact neutron dosimeter which can suppress a difference between an actual dose and a display dose. <P>SOLUTION: A neutron dosimeter includes: a mixed gas proportional counter tube 100, where<SP>3</SP>He and CH<SB>4</SB>are mixed and sealed at a prescribed mixture ratio; a separator 224 for separating detection signals from the mixed gas proportional counter tube for each wave height level; a counter 225 for counting detection signals for each wave height level separated by the separator; and a computing unit 227 for obtaining a dose by a count value counted by the counter. In the neutron dosimeter, coefficients used by the computing unit 227 are determined, based on a correlation among neutron detection sensitivity characteristics of<SP>3</SP>He, neutron detection sensitivity characteristics of CH<SB>4</SB>, and a dose equivalent at 1 cm depth. <P>COPYRIGHT: (C)2011,JPO&INPIT

Description

  The present invention includes a mixed gas proportional counter that detects neutron energy in various leaked neutron fields that may be exposed to neutrons, such as nuclear facilities and accelerator facilities, and processes detection signals output from the mixed gas proportional counter. The present invention relates to a neutron dosimeter that displays a dose equivalent at a depth of 1 cm, which is a practical amount.

  The effective dose equivalent, which is the degree of harmful effects on the human body due to neutron exposure, depends not only on the number of neutrons absorbed by the human body but also on the energy of the incident neutrons. The contribution ratio to the dose equivalent to the neutron energy is given by the International Commission on Radiological Protection (ICRP) in the form of a recommended conversion factor for neutron energy (MeV). In other words, since the effective dose indicating the degree of influence of neutron exposure on the human body is an amount that cannot be directly measured, a practical amount has been introduced in the ICRP 1990 recommendation as a countermeasure.

In the above-mentioned recommendation by ICRP, the concept of dose equivalent (H ^* (10)) at a depth of 1 cm is used. The evaluation of the probabilistic effects due to radiation exposure is based on the idea that it is appropriate that the dose equivalent considering this load is an effective dose equivalent because the degree of influence varies depending on the various tissues (organs) of the human body. ing. Since it is difficult to absolutely determine the effective dose equivalent under various conditions, the International Commission on Units and Measurement of Radiation (ICRU) has found that the 1 cm depth of the 30 cm diameter ICRU sphere phantom made of tissue equivalent material. Is adopted as the effective dose equivalent, and Sv (Sievert) is used as the unit. The above-mentioned ICRU sphere has a diameter of 30 made of a tissue equivalent material having a density of 1 g / cm ^3.
This is a spherical homogeneous phantom of cm, and the tissue equivalent material is composed of a mass composition of oxygen 76.2%, carbon 11.1%, hydrogen 10.1%, and nitrogen 2.6%.

The neutron detection principle is that the charged gas (BF ₃ , ³ He) is ionized by charged particles generated by the nuclear reaction between the incident thermal neutron and the sealed gas in the proportional counter to obtain a pulsed detection signal. It is. As for detection of neutrons using such a proportional counter, for example, a technique disclosed in Patent Document 1 (Japanese Patent Laid-Open No. 2008-14947) is well known. FIG. 12 is a diagram showing the relationship between the detection sensitivity characteristic of neutron energy by such a proportional counter and the dose equivalent (H ^* (10)) at a depth of 1 cm according to the ICRP74 recommendation. In the figure, the dotted line indicates the detection sensitivity of neutron energy in a detection device such as a proportional counter, and the solid line indicates the dose equivalent (H ^* (10)) at a depth of 1 cm according to the ICRP 74 recommendation. To calculate the dose indication value of the dose equivalent (H ^* (10)) at a depth of 1 cm, multiply the detection sensitivity of the detection device shown in the figure by the calibration constant (dose response) and calculate the total energy range. It is necessary to integrate with.
JP 2008-14947 A

However, in the conventional technique, in order to calculate the dose indication value of the 1 cm deep dose equivalent (H ^* (10)), the 1 cm deep dose equivalent (H ^* ) corresponding to the neutron energy of the detection device ^.
(10)) There is a problem that the response characteristics vary greatly depending on the dose equivalent (H ^* (10)) conversion coefficient in the 1 cm depth and the energy region of neutrons, and there is a high possibility that the dose will differ from the actual dose .

In order to solve the above problems, the invention according to claim 1 includes a mixed gas proportional counter tube in which ³ He and CH ₄ are mixed and sealed at a predetermined mixing ratio, and the mixed gas proportional counter tube. For classifying the detection signal from each wave height level, a counter for counting the detection signal for each wave height level sorted by the classifier, and multiplying the count value counted by the counter by a predetermined coefficient In the neutron dosimeter having the multiplier, the coefficient multiplied by the multiplier is determined based on the correlation between the neutron detection sensitivity characteristic of ³ He, the neutron detection sensitivity characteristic of CH ₄ , and the dose equivalent of 1 cm deep. It is characterized by that.

  The invention according to claim 2 is characterized in that, in the neutron dosimeter according to claim 1, Ar is enclosed in the mixed gas proportional counter.

According to the neutron dosimeter of the present invention, a calibration constant for neutron detection sensitivity of a single detection device is not used as in the prior art, but neutron detection sensitivity characteristics of ³ He, neutron detection sensitivity characteristics of CH ₄ , etc. Since the coefficient is determined based on the neutron detection sensitivity characteristics from a plurality of detection gases, it is possible to suppress the difference between the actual dose and the displayed dose.

In addition, according to the neutron dosimeter of the present invention, the dosimeter can be made compact by selecting a mixture of ³ He and CH ₄ as the combination of neutron detection gas sealed in the mixed gas proportional counter. Is possible.

Further, according to the neutron dosimeter of the present invention, since Ar is added to the mixed gas of ³ He and CH ₄ in the mixed gas proportional counter, the range of secondary reaction particles can be suppressed, and more accurate A detection signal can be obtained.

It is a figure which shows the typical cross section of the mixed gas proportional counter in the neutron dosimeter which concerns on embodiment of this invention. It is a figure which shows the outline | summary of the block structure of the neutron dosimeter which concerns on embodiment of this invention. It is a figure which shows the output in the thermal neutron field of the mixed gas proportional counter which concerns on this embodiment. It is a figure which shows the response test result (response with respect to 250 keV neutron) with respect to the acceleration neutron of the mixed gas proportional counter which concerns on this embodiment. It is a figure which shows the discrimination (in the case of the response with respect to 250 keV neutron) of the reaction in the mixed gas proportional counter output which concerns on this embodiment. It is a figure which shows the closed convex space by the response with respect to a monochromatic energy neutron. It is a figure which shows the neutron spectrum group represented by Formula (7). It is a figure which shows the calculation result of the response of the mixed gas proportional counter 100 with respect to a neutron spectrum. It is a figure which shows arrangement | positioning of the response with respect to a fission neutron spectrum. It is a figure which shows arrangement | positioning of the response with respect to the spectrum example of a fission neutron field. It is a figure which shows the output of the mixed gas proportional counter in a Kyoto University furnace. It is a figure which shows the relationship between the detection sensitivity characteristic of the neutron energy of a detection apparatus, and the dose equivalent (H ^* (10)) of 1 cm depth part by ICRP74 recommendation.

Hereinafter, embodiments of the neutron dosimeter of the present invention will be described with reference to the drawings. FIG. 1 is a diagram showing a schematic cross section of a mixed gas proportional counter in a neutron dosimeter according to an embodiment of the present invention, and FIG. 2 is a diagram showing an outline of a block configuration of the neutron dosimeter according to an embodiment of the present invention. It is.

  1 and 2, 100 is a mixed gas proportional counter, 110 is a moderator, 111 is a cylindrical metal body, 114 is an end plate, 115 is an insulator, 116 is a core wire, 117 is a filling gas, and 200 is a processing circuit. 221 is a power supply, 222 is a preamplifier, 223 is a waveform shaping amplifier, 224 is a multistage waveform separator, 225 is a counter, 226 is a multiplier, 227 is a calculator, and 228 is a display.

  A neutron dosimeter whose outline is shown in FIG. 2 detects neutrons in a power generation reactor, for example. Such a neutron dosimeter includes a mixed gas proportional counter 100 which is a neutron detector. This mixed gas proportional counter tube 100 is a cylindrical tube body, and its outer peripheral portion is formed of a cylindrical metal body 111 covered with a moderator 110.

  End plates 114 and 114 are provided at both ends of the cylindrical metal body 111, and an insulator 115 such as ceramic is provided at the center of each of the end plates 114. A cylindrical shape is provided between the insulators 115 and 115. A core wire 16 as an electrode is provided along the central axis of the metal body 110, and a filling gas 117 is enclosed in a space S defined by the cylindrical metal body 110 and the end plates 114 and 114.

  The cylindrical metal body 111 of the mixed gas proportional counter 100 is formed into a cylindrical shape with a conductive material such as stainless steel and functions as a cathode.

  A moderator 110 is provided on the outer surface of the cylindrical metal body 111. The neutron dosimeter receives an appropriate deceleration inside the moderator 110 of neutron contained in the radiation, and then reacts with a substance constituting the filling gas 117, and the reaction product ionizes the filling gas 117 to ion pairs. , And is accelerated by an electric field between the cylindrical metal body 110 (cathode) and the core wire 116 and is multiplied by a strong electric field in the vicinity of the core wire 116 to generate an electric signal, thereby making it possible to detect neutrons. .

  In this embodiment, a cylindrical shape is adopted as the shape of the mixed gas proportional counter 100. The reason for this will be described below. When the shape of the mixed gas proportional counter 100 is cylindrical, an appropriate output spectrum can be obtained, but there is a sensitivity difference depending on the incident direction of neutrons. On the other hand, when the mixed gas proportional counter is a spherical counter, the sensitivity can be made uniform in all directions, but the output spectrum is distorted by the internal electric field distribution. The distortion due to the electric field effect of the spherical counter output was confirmed with an actual machine. As a result, it was difficult to discriminate the reaction in the spherical counter with the required sensitivity due to the synergistic effect of spectral distortion and resolution degradation. Therefore, in the mixed gas proportional counter 100 according to the present embodiment, the cylindrical shape is selected as described above. The diameter of the cylinder of the mixed gas proportional counter 100 was set to 2.5 inches. The length of the tube was 7 inches to suppress the difference in sensitivity according to the neutron incident direction.

Next, the filling gas 117 sealed in the space S of the mixed gas proportional counter 100 will be described. In the present embodiment, ³ He and CH ₄ (methane) are mixed at a predetermined mixing ratio and sealed as a mixed gas sealed in the space S of the mixed gas proportional counter 100.

³ He gas exhibits a neutron absorption nuclear reaction, shows high sensitivity to thermal neutrons, and shows a tendency that detection sensitivity increases as the energy of neutron decreases.

On the other hand, hydrogen contained in methane gas undergoes a recoil reaction with neutrons, and a reaction to relatively high neutron energy is easy to detect. Instead of methane gas, it is possible to enclose either ethane or propane, or any combination of methane, ethane, or propane.

In the present embodiment, the result of investigation of combinations of gases for neutron detection encapsulated in mixed gas proportional counter 100 has been selected the mixing of ³ He and CH _4, The reason is as follows. The thermal neutron absorption reaction of ³ He forms a peak output spectrum with a large reaction cross section. CH ₄ detects a relatively fast neutron by the recoil reaction of hydrogen contained therein. The output of the recoil reaction forms a continuous spectrum and can be distinguished from the thermal neutron absorption reaction. In the present embodiment, Ar for suppressing the range of secondary reaction particles is added to the mixed gas of ³ He and CH ₄ sealed in the mixed gas proportional counter 100. The mixing ratio of the mixed gas of ³ He and CH ₄ was set so that a comparable coefficient could be obtained from both reactions for Am-Be neutrons. In consideration of suppression of other reactions, the composition ratio of ³ He and CH ₄ was set to about 1:30 to 1:60 in terms of gas partial pressure. Further, the total pressure of the gas sealed in the mixed gas proportional counter 100 was adjusted according to the size of the mixed gas proportional counter 100, and was set to 6 atm in consideration of dose sensitivity and the range of secondary particles.

Next, setting of conditions related to the moderator 111 used in the present embodiment will be described. ³ He and C
A thin moderator was provided in the mixed gas proportional counter 100 so as to balance the contribution of both H ₄ reactions to the dose evaluation, and the optimum land evaluation was performed by calculation. As a result, it was shown that a moderator of about 5 mm to 1 cm has a good balance of contribution to the dose of both reactions.

  Table 1 shows the specifications of the mixed gas proportional counter 100 according to the present embodiment.

以上のように構成される本実施形態に係る混合ガス比例計数管１００においては、約１００ｋｅＶを下回る低エネルギー中性子は、主に³Ｈｅ分子との衝突によって、³Ｈｅ（ｎ，ｐ）核反応により陽子ｐという荷電粒子を生成する。

In the mixed gas proportional counter 100 according to the present embodiment configured as described above, low energy neutrons below about 100 keV are mainly caused by collisions with ³ He molecules and by ³ He (n, p) nuclear reactions. A charged particle called proton p is generated.

Neutrons with energies of about 100 keV to about 10 MeV mainly generate H (n, n) p elastic scattering due to collision with CH ₄ molecules to generate charged particles called recoil protons p.

In the case of high-energy neutrons exceeding 10 MeV, H (n, n) in collision with CH ₄ molecules
In addition to recoil protons p by p elastic scattering, charged particles such as protons p and α particles by nuclear reactions such as C (n, p), C (n, α), and N (n, α) are also generated.

  These generated charged particles form ion pairs by ionization in the mixed gas proportional counter 100, and are amplified as a gas to output current pulses.

Next, an outline of the circuit configuration of the neutron dosimeter will be described. FIG. 2 is a block configuration diagram of the neutron dosimeter according to the embodiment of the present invention. A processing circuit unit 200 is connected to the mixed gas proportional counter 100 in the neutron dosimeter as shown in the block diagram of FIG. The processing circuit unit 200 includes a power source 221, a preamplifier 222, a waveform shaping amplifier 223, a multistage wave height discriminator 224, a counter 225, a multiplier 226, a calculator 227, and a display 228.

The high voltage power source 221 supplies a high voltage of 1000 V to 4000 V to the detector electrodes.
A detection signal based on a current output extracted from the electrode of the mixed gas proportional counter 100 of the neutron dosimeter is input to the preamplifier 222.

  The preamplifier 222 amplifies the detection signal until the wave height that can be used by the waveform shaping amplifier 223 is reached.

  The waveform shaping amplifier 223 outputs the detection signal amplified so that the detection signal has a wave height waveform that can be distinguished by the wave height discriminator 224. These preamplifier 222 and waveform shaping amplifier 223 constitute the amplifier of the present invention.

The wave height discriminator 224 has three stages of discrete levels, and outputs a detection signal of only a predetermined discrete level (from the lower wave height level to the upper wave height level) with respect to the input detection signal. To do. The first wave height discriminator 224-1 is less than the energy region of ³ He (n, p) nuclear reaction, the second wave height discriminator 224-2 is the energy region corresponding to ³ He (n, p) nuclear reaction, the third The wave height discriminator 224-3 is configured to output only a discretion level within a predetermined range above the energy region of ³ He (n, p) nuclear reaction.

  With this multistage wave height discriminator 224, unnecessary components below a certain level (noise signals and signals by γ rays) can be removed, and only desired detection signals by neutrons can be extracted. A plurality of detection signals classified according to wave height levels are output.

  The counter 225 inputs a plurality of detection signals classified according to wave height levels and outputs count values according to wave height levels. Since the detection signal is pulsed, the pulses are counted and a count value is output. The counter 225 includes a three-stage counter including a first counter 225-1, a second counter 225-2, and a third counter 225-3. Each of the first counter 225-1, the second counter 225-2, and the third counter 225-3 counts only a predetermined range of discrete levels.

The calculator 227 outputs a peripheral dose equivalent (1 cm dose equivalent) from the count value output from the counter 225. First, the count values of the first counter 225-1 and the third counter 225-3 are summed, and the ratio of the count value to the count value of the second counter 225-2 is obtained. Next, this is applied to a closed convex space for measuring the peripheral dose equivalent (1 cm dose equivalent), and the range of the conversion coefficient between the peripheral dose equivalent (1 cm dose equivalent) and the count value of the second counter 225-2 is obtained. This is multiplied by the count value of the second counter 225-2, and the range of the peripheral dose equivalent is output.

The display 228 displays the output from the calculator 227 so that it can be directly read as a range of ambient dose equivalent (depth 1 cm dose equivalent). The configuration described above is housed in a portable case. Such a display 227 enables a direct reading of the range of possible values of the peripheral dose equivalent (1 cm depth equivalent dose) with high accuracy.

According to the neutron dosimeter of the present invention as described above, the calibration constant for the neutron detection sensitivity of a single detection device is not used as in the prior art, but the neutron detection sensitivity characteristic of ³ He, the neutron detection of CH ₄ Since the coefficient is determined based on neutron detection sensitivity characteristics from a plurality of detection gases such as sensitivity characteristics, it is possible to suppress the difference between the actual dose and the displayed dose.

In addition, according to the neutron dosimeter of the present invention, the dosimeter can be made compact by selecting a mixture of ³ He and CH ₄ as a combination of neutron detection gas sealed in the mixed gas proportional counter 100. It becomes possible.

Further, according to the neutron dosimeter of the present invention, ³ He and C in the mixed gas proportional counter 100 are used.
Since Ar is added to the mixed gas of H _4, the range of secondary reaction particles can be suppressed, and a more accurate detection signal can be obtained.

  Next, a theoretical background of the neutron dosimeter according to the embodiment of the present invention configured as described above will be described. The neutron dosimeter according to the embodiment of the present invention is based on a dose evaluation method using a phase vector space.

  In general, the dose Q is expressed by a linear functional of the following equation (1).

ここで、ａは入射放射線の強度、ｑ（Ｅ）はＱをもたらすエネルギー依存の換算係数を示しており、φ（Ｅ）は規格化された入射放射線のエネルギースペクトルを示す。また、放射線検出器ｉの出力Ｎ_iも次式（２）のように表現することができる。

Here, a represents the intensity of incident radiation, q (E) represents an energy-dependent conversion factor that causes Q, and φ (E) represents the normalized energy spectrum of incident radiation. Further, the output N i of the radiation detector _i can also be expressed as the following equation (2).

ここで、ｇ_i（Ｅ）は測定量Ｎ_iをもたらす検出器ｉの応答関数である。
いま、放射線の強度ａの影響を相殺するために、ＱとＮ_jの比である一次分数汎関数Ｋと
、Ｎ_iとＮ_jの比である一次分数汎関数Ｒ_iを導入する。

Here, g _i (E) is the response function of the detector _i that gives the measured quantity N _i .
Now, in order to offset the effect of the intensity a of the radiation, the linear fractional functionals K is the ratio of Q and N _j, introducing N _i and N is the ratio of the _j linear fractional functionals R _i.

これらから、（_jＫ，_jＲ₁，_jＲ₂…_jＲ_j-1，_jＲ_j+1…_jＲ_i）を直交座標軸とする、ｉ次
元空間Ωを導入する。Ωは任意の放射線スペクトルφ（Ｅ）に対応する点ν（Ｅ）の集合で位相空間になる。また、Ωは閉集合であり、ベクトルν（Ｅ）について加法の可換則と結合則、分配則、スカラー乗法の結合則を満足しているため、Ωはベクトル空間でもある。すなわち、Ωは位相ベクトル空間である。

These, and _{(j K, j R 1,} j R 2 ... j R j-1, j R j + 1 ... j R i) the orthogonal coordinate axes, introducing i dimensional space Omega. Ω is a set of points ν (E) corresponding to an arbitrary radiation spectrum φ (E) and becomes a phase space. Further, Ω is a closed set, and Ω is also a vector space because the vector ν (E) satisfies the additive commutative law, the coupling law, the distribution law, and the scalar multiplication. That is, Ω is a phase vector space.

Here, since the coefficient a is canceled in the equations (3) and (4), the vector ν (aφ) corresponding to the radiation spectrum aφ (E) is represented by ν (E) corresponding to the radiation spectrum φ (E). )be equivalent to. Further, the vector v (φ _{1 + 2} ) corresponding to the combined spectrum c ₁ φ ₁ (E) + c ₂ φ ₂ (E) is expressed as the following equations (5) and (6).

これらから、点ｖ（φ₁₊₂）は、点ｖ（φ₁）と点ｖ（φ₂）を結ぶ線分上にある。さら
にこれを敷衍すると、あらゆるφ（Ｅ）が要素スペクトルφ_e（Ｅ）の線形合成で表せる
ならば、点ｖ（φ）の全ては要素スペクトルφ_e（Ｅ）に対応する点ｖ（φ_e）の集合で構成される閉凸空間ωに含まれることになる。φ_e（Ｅ）には単色エネルギースペクトルが
適用できるが、その他にも、線形合成で特定の中性子場を表現できる一連のスペクトルは要素スペクトルとなり得る。

From these, the point v (φ _{1 + 2} ) is on the line segment connecting the point v (φ ₁ ) and the point v (φ ₂ ). More Fuen this, if expressed by linear combination of any phi (E) is an element spectrum φ _{e (E),} the point v (phi) of all elements spectrum φ _{e (E)} corresponding points v (phi _e ) Are included in the closed convex space ω. A monochromatic energy spectrum can be applied to φ _e (E). In addition, a series of spectra that can express a specific neutron field by linear synthesis can be an element spectrum.

In this closed convex space ω, since a specific set of _j Ri can determine the (i−1) th condition, a chord in the K-axis direction is specified for ω. The range of this string is the range that _jK can take. In other words, the range of the conversion value for deriving the dose Q from the output N _{j of the} detector _j and the range of Q are determined by the ratio of the detector outputs. This is the multi-detector method.

  For future discussion, the phase vector space Ω is composed of dose response characteristics with respect to the target dose of multiple detectors, and will be referred to as “response space” below. In addition, the closed convex space ω formed for a specific spectrum is referred to as a “closed convex space”. This method can be applied to general dosimetry, but in this specification, application to evaluation of neutron dose is an embodiment.

  Next, a response test of the mixed gas proportional counter 100 in the present embodiment will be described. The above-described mixed gas proportional counter 100 was actually made and the response was confirmed. About this mixed gas proportional counter 100, the response characteristic test was done in several neutron fields, and the result is shown.

The response to thermal neutrons was confirmed with a graphite pile at the University of Tokyo. The result is shown in FIG. ³ H
The total energy peak of the thermal neutron absorption reaction of e was confirmed. Regarding the poor energy resolution, it was improved by adjusting the applied voltage in subsequent tests.

  Moreover, the response to fast neutrons and the discrimination performance of two types of reactions were confirmed using accelerator neutrons (JAEA JAERI FRS). An example of the result is shown in FIG.

  Although the resolution was somewhat poor, we were able to confirm the absorption spectrum of thermal neutrons and the continuous spectrum of recoiled protons with an appropriate maximum energy. Moreover, the spectrum of thermal neutron absorption reaction and the spectrum of recoil reaction could be distinguished by the baseline method. FIG. 5 shows the state of discrimination between both reactions.

Next, the closed convex space by the response to the monochromatic neutron will be described. The response characteristics of the mixed gas proportional counter 100 described above were evaluated in detail by calculation using NRESP-ANT, and a closed convex space was drawn in the response space. First, the monochromatic spectrum neutron was used as the element spectrum. The energy range was 1 × 10 ^{−3 to} 6.3 × 10 ⁸ eV, and the energy range was divided into 60 groups. The output spectrum of each reaction was extracted from the calculation results, and the closed convex space of the mixed gas proportional counter was drawn in the response space. In each case, the count of the absorption reaction of ³ He was used as a detector output (C _T ), and the count of elastic scattering of hydrogen was used as another detector output (C _F ). The evaluation target value was the ambient dose equivalent (H ^* (10)). The C _T as a common denominator, to form the response space. FIG. 6 shows the response space and the closed convex space. Due to the response to monochromatic spectrum neutrons, a closed convex space could be formed in the response space.

FIG. 6A is a closed convex space formed by the response to monochromatic energy neutrons formed in the response space of the mixed gas proportional counter 100, and each plot is a point (C _F / C _T , H / C _T ), the dotted line represents the boundary of the closed convex space. FIG. 6B is an enlarged view of FIG.

In FIG. 6B, when C _F / C _T = 10, the possible range of H ^* (10) / C _T is 0.002 to 5. This condition means that the possibility of values that can be taken as a dose conversion means has a two-digit range, which is not preferable for dose evaluation. This is because the closed convex space is obtained mathematically, and a physically inappropriate neutron energy spectrum is included in the evaluation range.

Next, the improvement of the closed convex section by specifying the neutron source will be described. In the previous section, a general closed convex space was drawn, but here, the evaluation target was narrowed down to a neutron field using fission neutrons as a radiation source. We applied the process of fission neutron deformation under the action of deceleration and amplification as an element spectrum.
As an example of the modification of the energy spectrum of fission neutrons, the transition of the in-reactor neutron spectrum F (u) was adopted. For F (u), the following general formula (7) shown by Kosako was adopted.

図７に、式（７）で導き出した中性子スペクトル群を示す。これらの中性子スペクトルに対する混合ガス比例計数管１００の応答をＮＲＥＳＰ−ＡＮＴで計算し、それらの結果をレスポンス空間中に配置したのが、図８である。図８（ａ）は、式（７）で表される中性子スペクトルを要素スペクトルとする閉凸空間であり、各プロットは図７の中性子スペクトルへの応答に対応する点である。また、閉凸空間の境界を実線で示している。点線は図６で示した閉凸空間であり、図８（ｂ）は図８（ａ）の拡大図である。

FIG. 7 shows the neutron spectrum group derived by the equation (7). FIG. 8 shows the response of the mixed gas proportional counter 100 to these neutron spectra calculated by NRESP-ANT, and the results are arranged in the response space. FIG. 8A is a closed convex space having the neutron spectrum represented by Equation (7) as an element spectrum, and each plot is a point corresponding to the response to the neutron spectrum of FIG. The boundary of the closed convex space is indicated by a solid line. A dotted line is the closed convex space shown in FIG. 6, and FIG. 8 (b) is an enlarged view of FIG. 8 (a).

The points corresponding to the energy spectrum given by Equation (7) were all located inside the closed convex space formed by the response to monochromatic neutrons. Furthermore, it can be confirmed that the closed convex space formed by the plot group corresponding to Expression (7) is greatly reduced. For example, when C _F / C _T = 10, the possible range of H ^* (10) / C _T is 0.02 to 0.16.

  The deformation of the fission neutron energy spectrum was also evaluated by another method. Using MCNP-5, 34 cases of neutron energy spectra in the process of softening fission neutron spectra by iron and concrete were obtained. FIG. 9 shows the result of evaluating the response of the counters to them similarly arranged in the response space. FIG. 9 (a) is the arrangement of the response to the slowed fission neutron spectrum calculated by MCNP. The dotted line indicates the boundary of the closed convex space shown in FIG. 6, and the solid line indicates the boundary of the reduced closed convex space shown in FIG. FIG. 9B is an enlarged view of FIG. All cases were shown to be in a reduced closed convex space. Moreover, the output set by the neutron spectrum calculated by MCNP-5 also forms a closed convex section, and the contraction is further advanced.

  Moreover, 85 data of the neutron spectrum caused by fission neutrons were referred to, responses in each were obtained, and the results were plotted in the response space (see FIG. 10). Most of the cases were shown to be inside the reduced closed convex space, with some exceptions. It was confirmed that any case that did not enter the contracted closed convex space deviated from the assumption of equation (7).

  FIG. 10 (a) is a diagram showing an arrangement of responses to a spectrum example of a fission neutron field. A dotted line shows the boundary of the closed convex space shown in FIG. 6, and a solid line shows the boundary of the reduced closed convex space shown in FIG. FIG. 10B is an enlarged view of FIG.

From the above, H ^* (10) / C _T is limited based on the value of C _F / C _T , and therefore H ^* (1
The range that 0) can take was determined.

  Next, the dose evaluation will be examined. Dose evaluation using a closed convex space has provided a range of possible dose values, but it is considered that it is required to determine one dose value practically. The closed convex space in this measurement has an arcuate shape, and it is considered that the measurement results are distributed in the interior. As a representative value, it is appropriate to use the portion of the string that forms the upper side of the closed convex space.

First, a response test using fission neutrons will be described. Actually, the response test of the mixed gas proportional counter was conducted in several fission neutron fields. FIG. 11 shows the experimental results at the Kyoto University furnace. Based on the neutron spectrum measurement result at the measurement point, the dose conversion coefficient specific to the site is evaluated, and the relationship between the values of H / C _T and C _F / C _T is shown in Table 2. Dose conversion coefficient and H / C _T value seem to show correlation

DESCRIPTION OF SYMBOLS 100 ... Mixed gas proportional counter, 110 ... Moderator, 111 ... Cylindrical metal body, 114 ... End plate, 115 ... Insulator, 116 ... Core wire, 117 ... Filling gas, 200 ... processing circuit unit, 221 ... power supply, 222 ... preamplifier, 223 ... waveform shaping amplifier, 224 ... multi-stage waveform separator, 225 ... counter, 226,.・ Multiplier, 227 ... Calculator, 228 ... Display

Claims (2)

A mixed gas proportional counter in which ³ He and CH ₄ are mixed and sealed at a predetermined mixing ratio;
A separator for separating the detection signal from the mixed gas proportional counter for each wave height level;
A counter that counts a detection signal for each wave height level sorted by the classifier;
In a neutron dosimeter having a calculator that calculates a dose using the count value counted by the counter,
The coefficient used in the computing unit is determined based on a correlation between a neutron detection sensitivity characteristic of ³ He, a neutron detection sensitivity characteristic of CH ₄ , and a dose equivalent at a depth of 1 cm. The neutron dosimeter according to claim 1, wherein Ar is enclosed in the mixed gas proportional counter.

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