JP2635860B2 - Radioactivity evaluation method for solidified radioactive waste - Google Patents
Radioactivity evaluation method for solidified radioactive wasteInfo
- Publication number
- JP2635860B2 JP2635860B2 JP17953191A JP17953191A JP2635860B2 JP 2635860 B2 JP2635860 B2 JP 2635860B2 JP 17953191 A JP17953191 A JP 17953191A JP 17953191 A JP17953191 A JP 17953191A JP 2635860 B2 JP2635860 B2 JP 2635860B2
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- JP
- Japan
- Prior art keywords
- radioactivity
- detection efficiency
- vitrified
- measurement
- radioactive waste
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Description
【0001】[0001]
【産業上の利用分野】本発明は、放射性廃棄物固化体の
放射能評価方法に係り、特に、原子力発電施設で使用し
た核燃料物質の再処理等において発生する高レベル放射
性ガラス固化体に含まれる放射性核種の放射能を非破壊
測定にて絶対値定量する放射性廃棄物固化体の放射能評
価方法に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for evaluating the radioactivity of a solidified radioactive waste, and more particularly to a high-level radioactive solidified material generated in the reprocessing of nuclear fuel materials used in nuclear power plants. The present invention relates to a radioactivity evaluation method for solidified radioactive waste, in which the absolute value of radioactivity of radionuclides is determined by nondestructive measurement.
【0002】[0002]
【従来の技術】一般に、核燃料サイクル施設等で取り扱
われる放射性廃棄物固化体の放射能を測定、評価するこ
とは、安全評価等の上から重要である。2. Description of the Related Art Generally, it is important from the viewpoint of safety evaluation and the like to measure and evaluate the radioactivity of a solidified radioactive waste handled in a nuclear fuel cycle facility or the like.
【0003】このような放射性廃棄物固化体の一つとし
て、原子力発電施設で使用した核燃料物質の再処理等に
おいて発生する高レベル放射性ガラス固化体がある。と
ころが、従来このような高レベル放射性ガラス固化体中
の放射性核種の放射能を非破壊測定にて絶対値定量する
技術はなく、ガラス固化する以前に、直接γ線、α線中
性子線等を測定することにより、核種別放射能を定量す
る方法、あるいは、線量率測定器によりガラス固化体の
放射線量率のみを測定する方法等があるだけであった。
なお、濃度既知のRIを含む試験用小型ガラス固化体の
試験測定例はある。[0003] As one of such solidified radioactive wastes, there is a high-level radioactive solidified solid generated in the reprocessing of nuclear fuel materials used in nuclear power plants. However, there is no conventional technology for non-destructive measurement of the radioactivity of radioactive nuclides in such high-level radioactive vitrified solids.Before vitrification, directly measure γ-rays, α-ray neutrons, etc. By doing so, there is only a method of quantifying the radioactivity of the nuclide, or a method of measuring only the radiation dose rate of the vitrified material using a dose rate measuring device.
In addition, there is a test measurement example of a small test solidified glass containing RI with a known concentration.
【0004】[0004]
【発明が解決しようとする課題】上述したように、従来
は、ガラス固化する以前においては、直接γ線、α線、
中性子線等を測定し核種別放射能を定量することができ
るものの、一旦ガラス固化された高放射性廃棄物の核種
別放射能を定量するためには、再び破壊して放射能を分
析測定しなければならず、作業能率、作業員の被曝管理
等の点から実際上行うことはできない。As described above, conventionally, prior to vitrification, direct γ-rays, α-rays,
Although it is possible to quantify radioactivity by measuring neutron beams, etc., in order to quantify radioactivity of vitrified radioactive waste, it must be destroyed again and analyzed for radioactivity. In practice, it cannot be performed in terms of work efficiency and worker exposure control.
【0005】また、上述したような放射線量率の測定で
は、放射性核種の絶対値定量はできない。このため、放
射性廃棄物の処分においては、処分後の安全評価上、放
射性核種毎の放射能あるいは放射能濃度を把握すること
が好ましいが、このような評価は従来行われていなかっ
た。[0005] In the measurement of the radiation dose rate as described above, the absolute value of the radionuclide cannot be determined. For this reason, in the disposal of radioactive waste, it is preferable to grasp the radioactivity or radioactivity concentration for each radionuclide for safety evaluation after disposal, but such evaluation has not been performed so far.
【0006】本発明は、かかる従来の事情に対処してな
されたもので、放射性廃棄物固化体に含まれる放射性核
種毎の放射能を非破壊測定にて容易に絶対値定量するこ
とができ、従来に較べて安全性の向上を図ることのでき
る放射性廃棄物固化体の放射能評価方法を提供しようと
するものである。The present invention has been made in view of such conventional circumstances, and the radioactivity of each radionuclide contained in the solidified radioactive waste can be easily quantified in absolute value by nondestructive measurement. An object of the present invention is to provide a method for evaluating the radioactivity of a solidified radioactive waste, which can improve the safety as compared with the related art.
【0007】[0007]
【課題を解決するための手段】すなわち本発明は、内部
に放射能濃度既知のRI線源を配置可能に構成され、か
つ、このRI線源の位置を複数の箇所に移動可能に構成
された放射能校正用模擬供試体を回転させつつ放出γ線
の計測を行い、放射性核種に対するγ線の検出効率を求
める工程を、前記放射能校正用模擬供試体内の前記RI
線源の位置を変えながら繰り返し行って検出効率分布を
求め、この検出効率分布から単位体積あたりの平均的な
放射能検出効率を導出し、この後、評価対象の放射性廃
棄物固化体のγ線の計測を行い、この計測結果を前記単
位体積あたりの平均的な放射能検出効率によって補正す
ることを特徴とする。That is, the present invention is configured such that an RI source having a known radioactivity concentration can be disposed therein, and that the position of the RI source can be moved to a plurality of locations. The step of measuring emission γ-rays while rotating the simulation sample for radioactivity calibration and obtaining the detection efficiency of γ-rays for radioactive nuclides comprises the step of measuring the RI in the simulation sample for radioactivity calibration.
The detection efficiency distribution is repeatedly obtained while changing the position of the radiation source to derive an average radioactivity detection efficiency per unit volume from the detection efficiency distribution, and thereafter, γ-rays of the solidified radioactive waste to be evaluated are obtained. , And the measurement result is corrected by the average radioactivity detection efficiency per unit volume.
【0008】[0008]
【作用】上記構成の本発明の放射性廃棄物固化体の放射
能評価方法では、内部に放射能濃度既知のRI線源を配
置可能に構成され、かつ、このRI線源の位置を複数の
箇所に移動可能に構成された放射能校正用模擬供試体を
用いる。すなわち、例えばガラス固化体状の放射線源か
ら放出されるγ線の検出効率を3次元的に測定できるよ
うな放射能校正用模擬線源として、Cs−137やCo
−60等のRI線源を分布配置できる模擬ガラス固化体
を用い、高レベル放射性ガラス固化体の放射能測定と同
一測定条件にて、RI線源の配置を変えて模擬ガラス固
化体からの放射能測定を繰り返すことにより、ガラス固
化体の単位体積あたりの平均的な放射能の検出効率を求
める。According to the radioactivity evaluation method for a solidified radioactive waste of the present invention having the above-described configuration, an RI source having a known radioactivity concentration can be disposed inside the source, and the position of the RI source can be determined at a plurality of locations. Use a simulated specimen for calibration of radioactivity that is configured to be movable. That is, Cs-137 or Cos-137 is used as a simulated radiation source for radioactivity calibration so that the detection efficiency of gamma rays emitted from a vitrified radiation source can be measured three-dimensionally.
Using a simulated vitrified body capable of distributing and arranging RI radiation sources such as -60, radiation from the simulated vitrified body by changing the arrangement of the RI source under the same measurement conditions as the radioactivity measurement of the high-level radioactive vitrified body By repeating the measurement of the radioactivity, the average detection efficiency of radioactivity per unit volume of the vitrified product is determined.
【0009】つまり、例えば、まず校正用模擬ガラス固
化体の任意の半径方向位置に、RI線源を配置し、高さ
方向の測定位置を固定した状態でその位置での検出効率
測定をRI線源の半径方向位置を変えて繰り返すことに
より、半径方向の検出効率分布を求める。次に、この半
径方向検出効率分布の測定を高さ方向の測定位置を、細
かく変えて繰り返し測定することで高さ方向の検出効率
分布を得ることができ、結果としてガラス固化体の立体
角的な検出効率分布を求め、全体を平均することによ
り、検出効率の分布を考慮した単位体積あたりの平均的
な検出効率が得られる。That is, for example, first, an RI ray source is arranged at an arbitrary radial position of the simulated vitrified body for calibration, and the measurement of the detection efficiency at that position is performed while the measurement position in the height direction is fixed. By repeatedly changing the position of the source in the radial direction, the distribution of detection efficiency in the radial direction is obtained. Next, the measurement efficiency distribution in the radial direction can be obtained by repeatedly measuring the measurement position in the height direction with the measurement position in the height direction being finely changed to obtain the detection efficiency distribution in the height direction. An average detection efficiency per unit volume in consideration of the distribution of the detection efficiencies can be obtained by obtaining the optimum detection efficiency distribution and averaging the entire distribution.
【0010】そして、この検出効率と評価対象核種のγ
線の放出率を用いて、高レベル放射性ガラス固化体の外
部からのγ線スペクトル測定による計数値を補正するこ
とにより、高レベル放射性ガラス固化体全体に含まれる
放射性核種の放射能を定量する。The detection efficiency and the γ of the nuclide to be evaluated
The radioactivity of the radionuclide contained in the entire high-level radioactive vitrified matter is quantified by correcting the count value obtained by measuring the gamma ray spectrum from the outside of the high-level radioactive vitrified substance using the radiation rate of the radiation.
【0011】この方法を用いれば、例えば大型であって
製作及び取扱いが難しい高レベル放射性固化体と同一形
状の校正用ガラス固化体を用いることなく、放射能測定
の校正を行うことができ、かつ、平均的な検出効率を求
めておくので、高レベル放射性ガラス固化体の測定は、
任意箇所の外部からのγ線スペクトル測定を行うだけで
よい。By using this method, it is possible to calibrate radioactivity measurement without using, for example, a large-sized high-level radioactive solidified body which is difficult to manufacture and handle and having the same shape as a calibrated glass solidified body. In order to determine the average detection efficiency, measurement of high-level radioactive glass
It is only necessary to measure the gamma-ray spectrum from an external portion at any position.
【0012】[0012]
【実施例】以下、本発明の放射性廃棄物固化体の放射能
評価方法の詳細を図面を参照して実施例について説明す
る。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The details of the radioactivity evaluation method for solidified radioactive waste of the present invention will be described below with reference to the drawings.
【0013】図1は、本発明の一実施例における放射能
測定システムの構成を示すもので、図において1は、高
レベル放射性ガラス固化体を示している。この高レベル
放射性ガラス固化体1は、上下動および回動自在に構成
されたガラス固化体駆動装置2上に載置され、この高レ
ベル放射性ガラス固化体1から放出されるγ線は、放射
線遮蔽壁3を貫通するように設置されたγ線コリメータ
4を介して放射線検出器5で計数するよう構成されてい
る。FIG. 1 shows a configuration of a radioactivity measuring system according to an embodiment of the present invention. In FIG. 1, reference numeral 1 denotes a high-level radioactive vitrified body. The high-level radioactive vitrified body 1 is mounted on a vitrified body driving device 2 configured to be vertically movable and rotatable, and γ-rays emitted from the high-level radioactive vitrified body 1 are shielded from radiation. The radiation detector 5 is configured to count through a γ-ray collimator 4 installed so as to penetrate the wall 3.
【0014】一方、図2において、6はRI線源を装荷
した絶対値校正用の模擬ガラス固化体を示している。こ
の模擬ガラス固化体6は、評価対象である高レベル放射
性ガラス固化体1と同様な材質によってほぼ同径の円柱
状に構成されているが、高さ方向の長さは、高レベル放
射性ガラス固化体1に較べて短く、小型に構成されてい
る。また、この模擬ガラス固化体6の内部には、所望の
RI線源を装荷可能に構成されており、かつ、このRI
線源を径方向の複数の異なる位置に配置することができ
るよう構成されている。なお、本実施例では、後述する
図5にも示すように、模擬ガラス固化体6の中心から、
径方向外周までの11か所にRI線源を配置することがで
きるよう構成されている。On the other hand, in FIG. 2, reference numeral 6 denotes a simulated vitrified body for absolute value calibration loaded with an RI source. The simulated vitrified body 6 is made of a material similar to that of the high-level radioactive vitrified body 1 to be evaluated, and is formed in a columnar shape having substantially the same diameter. It is shorter and smaller than the body 1. Further, inside the simulated vitrified body 6, a desired RI radiation source is configured to be loaded, and the RI
It is configured such that the radiation source can be arranged at a plurality of different positions in the radial direction. In this example, as shown in FIG. 5 described later, from the center of the simulated vitrified body 6,
It is configured so that RI radiation sources can be arranged at eleven locations up to the radial outer periphery.
【0015】本実施例では、高レベル放射性ガラス固化
体1の放射能測定に先立ち、まず、ガラス固化体駆動装
置2上に模擬ガラス固化体6を載置し、この模擬ガラス
固化体6を回転させながら、放射線検出器5で放射線の
計測を実施する。In this embodiment, prior to the measurement of the radioactivity of the high-level radioactive vitrified body 1, first, the simulated vitrified body 6 is placed on the vitrified body driving device 2, and the simulated vitrified body 6 is rotated. While performing the measurement, the radiation detector 5 measures the radiation.
【0016】図3は、本実施例における放射能測定評価
の手順を示すもので、例えば、高レベル放射性ガラス固
化体中のCs−137の放射能を評価する場合、以下の
手順にて評価を行う。FIG. 3 shows the procedure of radioactivity measurement and evaluation in the present embodiment. For example, when radioactivity of Cs-137 in a high-level radioactive vitrified substance is evaluated, the evaluation is performed according to the following procedure. Do.
【0017】すなわち、高レベル放射性ガラス固化体1
の放射能測定に先立ち、まず濃度既知のCs−137線
源を任意の半径方向位置に装荷した模擬ガラス固化体6
のγ線スペクトル測定を行う(100)。That is, the high-level radioactive vitrified product 1
Prior to the measurement of the radioactivity, a simulated vitrified body 6 in which a Cs-137 source having a known concentration was loaded at an arbitrary radial position.
(100).
【0018】この測定としては、まず、模擬ガラス固化
体6をガラス固化体駆動装置2上に載置し、ある高さ位
置に設定した状態で、模擬ガラス固化体6を回転させな
がら放射線検出器5でγ線スペクトル測定を行うことに
より、ある高さ位置における回転平均計数率を測定する
(101)。In this measurement, first, the simulated vitrified body 6 is placed on the vitrified body driving device 2 and set at a certain height position while rotating the simulated vitrified body 6 while detecting the radiation detector. The rotational average count rate at a certain height position is measured by performing γ-ray spectrum measurement in step 5 (101).
【0019】次に、模擬ガラス固化体6を高さ方向に若
干移動し(102)、再び回転平均計数率を測定する
(101)。Next, the simulated vitrified body 6 is slightly moved in the height direction (102), and the rotational average count rate is measured again (101).
【0020】この高さ方向への移動と回転測定を繰り返
すことにより、図4に示すような任意半径方向位置にお
ける計数率の高さ方向の分布を得る(103)。なお、
図4のグラフは、γ線コリメータの中心を基準にした高
さ方向位置に対し、その高さ方向位置での計数率をプロ
ットしたものである。By repeating the movement in the height direction and the rotation measurement, a distribution in the height direction of the count rate at an arbitrary radial position as shown in FIG. 4 is obtained (103). In addition,
The graph of FIG. 4 plots the counting rate at the height position with respect to the height position based on the center of the γ-ray collimator.
【0021】次に、図4の斜線部分の面積より高さ方向
平均の計数率を求め、その高さ方向平均の計数率を、C
s−137線源の単位時間あたりのγ線発生量とその半
径位置でのγ線コリメータから挑むことができる面積で
除することで、任意半径方向位置における単位面積あた
りの検出効率Ehを得る(104)。Next, the average count rate in the height direction is obtained from the area of the hatched portion in FIG.
By dividing the γ-ray generation amount per unit time of the s-137 source by the area that can be challenged by the γ-ray collimator at that radial position, the detection efficiency Eh per unit area at an arbitrary radial position is obtained ( 104).
【0022】この後、Cs−137線源の半径方向の位
置を変え(105)、上記101から104の測定およ
び解析を繰り返し、図5に示すような半径方向の検出効
率分布を得る(106)。Thereafter, the position of the Cs-137 source in the radial direction is changed (105), and the measurement and analysis of 101 to 104 are repeated to obtain a radial detection efficiency distribution as shown in FIG. 5 (106). .
【0023】図5のグラフは、模擬ガラス固化体6の円
中心空の半径方向距離に対して、各半径方向距離におけ
る検出効率Ehに、円周方向の体積の重みを乗じた値を
プロットしたものである。そして、次にこの斜線部の面
積から、模擬ガラス固化体6中のCs−137から放出
されるγ線に対する単位体積あたりの平均的な検出効率
Eを求める(107)。The graph of FIG. 5 plots the value obtained by multiplying the detection efficiency Eh at each radial distance by the weight of the volume in the circumferential direction with respect to the radial distance in the center of the circle of the simulated vitrified body 6. Things. Then, an average detection efficiency E per unit volume for γ-rays emitted from Cs-137 in the simulated vitrified solid 6 is determined from the area of the hatched portion (107).
【0024】次に、図1に示したように、高レベル放射
性ガラス固化体1のγ線スペクトル測定を実施する(2
00)。Next, as shown in FIG. 1, γ-ray spectrum measurement of the high-level radioactive vitrified product 1 is performed (2).
00).
【0025】すなわち、まず、高さ方向の任意静止位置
での高レベル放射性ガラス固化体1のγ線スペクトル測
定により、Cs−137から放出される662KeV のエネ
ルギーを持つγ線のガラス固化体単位体積あたりの計数
率Cを測定する(201)。そして、ガラス固化体中の
放射能は均質に分布していると仮定し(202)、この
計数率Cを、Cs−137の662 KeV のγ線放出率Bで
除し(203)、さらに上述した模擬ガラス固化体6の
測定から得た検出効率Eで除することにより、高レベル
放射性ガラス固化体中1の単位体積中に含まれるCs−
137の放射能Avを得る(204)。That is, first, a γ-ray vitrified unit volume of γ-ray having an energy of 662 KeV emitted from Cs-137 is measured by a γ-ray spectrum measurement of the high-level radioactive vitrified body 1 at an arbitrary rest position in the height direction. The counting rate C per unit is measured (201). Then, it is assumed that the radioactivity in the vitrified material is homogeneously distributed (202), and this count rate C is divided by the γ-ray emission rate B of 662 KeV of Cs-137 (203). By dividing by the detection efficiency E obtained from the measurement of the simulated vitrified material 6, Cs- contained in one unit volume of the high-level radioactive vitrified material
137 radioactivity Av is obtained (204).
【0026】そして、このようにして求めた放射能Av
に、ガラス固化体の体積Vを乗ずることにより(20
5)、高レベル放射性ガラス固化体1中のCs−137
の全放射能を求める(206)。The radioactivity Av thus obtained is
Is multiplied by the volume V of the vitrified product (20
5), Cs-137 in the high-level radioactive vitrified product 1
Is determined (206).
【0027】このようにして、本実施例では、簡易な校
正用の模擬ガラス固化体6を用いることにより、簡単な
校正作業で放射能評価を行うことができる。なお、他の
核種についても、上述したCs−137の場合と同様に
放射能評価を行うことができる。As described above, in this embodiment, by using the simulated vitrified body 6 for simple calibration, the radioactivity can be evaluated by a simple calibration work. In addition, radioactivity evaluation can be performed for other nuclides in the same manner as in the case of Cs-137 described above.
【0028】次に、高レベル放射性ガラス固化体1中に
含まれる任意の放射性核種の放射能評価を行う場合につ
いて、図6を参照して説明する。Next, a case of evaluating the radioactivity of an arbitrary radionuclide contained in the high-level radioactive vitrified product 1 will be described with reference to FIG.
【0029】まず、濃度既知のエネルギーεのγ線を放
出するRI線源を装荷した模擬ガラス固化体6のγ線ス
ペクトル測定(300)では、前述した手順によって測
定を行い、エネルギーεのγ線に対するガラス固化体の
単位体積あたりの平均的な検出効率E(ε)を求める
(301)。First, in the γ-ray spectrum measurement (300) of the simulated vitrified body 6 loaded with an RI ray source that emits γ-rays of energy ε of known concentration, the measurement is performed according to the above-described procedure, and γ-rays of energy ε are measured. The average detection efficiency E (ε) per unit volume of the vitrified body with respect to is determined (301).
【0030】次に、エネルギーεの異なるRI線源に変
更して(302)、同様に検出効率E(ε)をめる(3
02)。このような工程を複数回繰り返して行い、得ら
れた複数のE(ε)より、検出効率関数f(ε)を得る
(303)。Next, an RI source having a different energy ε is changed (302), and the detection efficiency E (ε) is similarly increased (3).
02). Such a process is repeated a plurality of times, and a detection efficiency function f (ε) is obtained from the obtained plurality of E (ε) (303).
【0031】一方、高レベル放射性ガラス固化体1のγ
線スペクトル測定(400)においては、まず、γ線ス
ペクトル測定によって任意核種の計数率C(ε)を求め
(401)、その核種のγ線エネルギーに対応する検出
効率を検出効率関数f(ε)から導出して、前述した実
施例の場合と同様にして、高レベル放射性ガラス固化体
中1に含まれる任意核種の単位体積あたりの放射能Av
(ε)を求める(402)。On the other hand, γ of the high level radioactive vitrified material 1
In the line spectrum measurement (400), first, the counting rate C (ε) of an arbitrary nuclide is determined by gamma ray spectrum measurement (401), and the detection efficiency corresponding to the γ-ray energy of the nuclide is determined by a detection efficiency function f (ε). And the radioactivity Av per unit volume of an arbitrary nuclide contained in 1 in the high-level radioactive vitrified material in the same manner as in the above-described embodiment.
(Ε) is obtained (402).
【0032】そして、ガラス固化体の体積から、任意核
種の放射能A(ε)を求める(403)。Then, the radioactivity A (ε) of an arbitrary nuclide is determined from the volume of the vitrified material (403).
【0033】この実施例では、検出効率を関数化するた
め、高レベル放射性廃棄物を取り扱う施設のような、構
造及び管理体制上作業性が比較的劣る施設以外で、検出
効率の測定作業を行うことができる。このため、作業性
が向上し、放射能評価作業の効率が向上する。In this embodiment, in order to make the detection efficiency a function, measurement work of the detection efficiency is performed at a facility other than a facility having relatively low workability due to its structure and management system, such as a facility handling high-level radioactive waste. be able to. Therefore, the workability is improved, and the efficiency of the radioactivity evaluation work is improved.
【0034】次に、図7および図8を参照して他の実施
例について説明する。Next, another embodiment will be described with reference to FIGS.
【0035】図7は、高レベル放射性ガラス固化体1の
スキャン測定範囲を示すもので、この実施例では、高レ
ベル放射性ガラス固化体1の高さ方向のスキャン範囲
は、高レベル放射性ガラス固化体1の高さより長い範囲
に設定する。このため、例えばCs−137の662 KeV
のγ線の計数は同図の斜線部分30のようになる。FIG. 7 shows a scan measurement range of the high-level radioactive vitrified body 1. In this embodiment, the scanning range in the height direction of the high-level radioactive vitrified body 1 is the high-level radioactive vitrified body. Set to a range longer than the height of 1. Therefore, for example, 662 KeV of Cs-137
The count of γ-rays is as shown by the shaded portion 30 in FIG.
【0036】本実施例では、図8に示すように、高レベ
ル放射性ガラス固化体1のγ線スペクトル測定において
(500)、まず、スキャン範囲内の平均計数率Cave
を、上記斜線部分30の面積をスキャン時間Tで除した
値として求め(501)、このCave 30を、Cs−1
37に対する単位体積あたりの検出効率Eで除し(50
2)、Cs−137のγ線放出率Bで除することにより
(503)、高レベル放射性ガラス固化体の単位体積あ
たりに含まれるCs−137の放射能Avを求める(5
04)。In this embodiment, as shown in FIG. 8, in the measurement of the γ-ray spectrum of the high-level radioactive vitrified product 1 (500), first, the average count rate Cave within the scan range is obtained.
Is obtained as a value obtained by dividing the area of the shaded portion 30 by the scan time T (501), and this Cave 30 is calculated as Cs-1.
37 divided by the detection efficiency E per unit volume (50
2) By dividing by γ-ray emission rate B of Cs-137 (503), radioactivity Av of Cs-137 contained per unit volume of the high-level radioactive vitrified product is obtained (5).
04).
【0037】そして、この単位体積あたりの放射能Av
にスキャン体積Vsca を乗ずることにより(505)、
高レベル放射性ガラス固化体中のCs−137の放射能
Aを求める(506)。The radioactivity Av per unit volume
Multiplied by the scan volume Vsca (505),
The radioactivity A of Cs-137 in the high-level radioactive vitrified matter is determined (506).
【0038】本実施例によれば、高レベル放射性ガラス
固化体1の体積が不明な場合でも、放射能の絶対値定量
を行うことができる。According to this embodiment, the absolute value of radioactivity can be determined even when the volume of the high-level radioactive vitrified product 1 is unknown.
【0039】[0039]
【発明の効果】以上説明したように、本発明によれば、
放射性廃棄物固化体に含まれる放射性核種毎の放射能を
非破壊測定にて容易に絶対値定量することができ、従来
に較べて安全性の向上を図ることができる。As described above, according to the present invention,
The absolute value of the radioactivity of each radionuclide contained in the solidified radioactive waste can be easily determined by nondestructive measurement, and safety can be improved as compared with the conventional method.
【図1】本発明の一実施例における測定システムの構成
を示す図。FIG. 1 is a diagram showing a configuration of a measurement system according to an embodiment of the present invention.
【図2】本発明の一実施例における測定システムの構成
を示す図。FIG. 2 is a diagram showing a configuration of a measurement system according to an embodiment of the present invention.
【図3】本発明の一実施例における放射能測定評価手順
を示す図。FIG. 3 is a diagram showing a procedure for measuring and evaluating radioactivity in one embodiment of the present invention.
【図4】本発明の一実施例の放射能校正における計数率
分布を示す図。FIG. 4 is a diagram showing a count rate distribution in radioactivity calibration according to one embodiment of the present invention.
【図5】本発明の一実施例の放射能校正における検出効
率分布を示す図。FIG. 5 is a diagram showing a detection efficiency distribution in radioactivity calibration according to one embodiment of the present invention.
【図6】他の実施例における放射能測定評価手順を示す
図。FIG. 6 is a diagram showing a radioactivity measurement evaluation procedure in another embodiment.
【図7】他の実施例におけるスキャン範囲を説明するた
めの図。FIG. 7 is a diagram for explaining a scan range in another embodiment.
【図8】他の実施例における放射能測定評価手順を示す
図。FIG. 8 is a diagram showing a radioactivity measurement evaluation procedure in another example.
1 高レベル放射性ガラス固化体 2 ガラス固化体駆動装置 3 放射線遮蔽壁 4 γ線コリメータ 5 放射線検出器 6 校正用の模擬ガラス固化体 REFERENCE SIGNS LIST 1 high-level radioactive vitrified substance 2 vitrified substance driving device 3 radiation shielding wall 4 γ-ray collimator 5 radiation detector 6 simulated vitrified substance for calibration
Claims (1)
可能に構成され、かつ、このRI線源の位置を複数の箇
所に移動可能に構成された放射能校正用模擬供試体を回
転させつつ放出γ線の計測を行い、放射性核種に対する
γ線の検出効率を求める工程を、前記放射能校正用模擬
供試体内の前記RI線源の位置を変えながら繰り返し行
って検出効率分布を求め、この検出効率分布から単位体
積あたりの平均的な放射能検出効率を導出し、 この後、評価対象の放射性廃棄物固化体のγ線の計測を
行い、この計測結果を前記単位体積あたりの平均的な放
射能検出効率によって補正することを特徴とする放射性
廃棄物固化体の放射能評価方法。A simulated specimen for calibration of radioactivity in which an RI source having a known radioactivity concentration can be arranged and the position of the RI source can be moved to a plurality of locations. Measuring the emission γ-rays while performing the process of obtaining the detection efficiency of γ-rays for radioactive nuclides, while repeatedly changing the position of the RI source in the simulation sample for radioactivity calibration to obtain the detection efficiency distribution The average radioactivity detection efficiency per unit volume is derived from the detection efficiency distribution, and thereafter, the γ-ray of the radioactive waste solid to be evaluated is measured, and the measurement result is averaged per unit volume. A method for evaluating the radioactivity of a solidified radioactive waste, wherein the radioactive waste is corrected by a specific radioactivity detection efficiency.
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JP17953191A JP2635860B2 (en) | 1991-07-19 | 1991-07-19 | Radioactivity evaluation method for solidified radioactive waste |
Applications Claiming Priority (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP17953191A JP2635860B2 (en) | 1991-07-19 | 1991-07-19 | Radioactivity evaluation method for solidified radioactive waste |
Publications (2)
Publication Number | Publication Date |
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JPH0527044A JPH0527044A (en) | 1993-02-05 |
JP2635860B2 true JP2635860B2 (en) | 1997-07-30 |
Family
ID=16067387
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JP17953191A Expired - Fee Related JP2635860B2 (en) | 1991-07-19 | 1991-07-19 | Radioactivity evaluation method for solidified radioactive waste |
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JP (1) | JP2635860B2 (en) |
Cited By (1)
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KR102025555B1 (en) * | 2018-04-19 | 2019-09-26 | 포항공과대학교 산학협력단 | A homogeneity test method of solidified radioactive waste and its apparatus |
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JP4568818B2 (en) * | 2005-09-29 | 2010-10-27 | 独立行政法人 日本原子力研究開発機構 | Visualization device using gamma ray source |
JP5337288B1 (en) * | 2012-09-25 | 2013-11-06 | 有限会社品川通信計装サービス | Radioactivity measuring instrument measurement accuracy ensuring confirmation acquisition method for food radioactivity measurement and radioactivity measuring instrument measurement accuracy ensuring confirmation acquisition device for food |
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1991
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KR102025555B1 (en) * | 2018-04-19 | 2019-09-26 | 포항공과대학교 산학협력단 | A homogeneity test method of solidified radioactive waste and its apparatus |
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