JPH0659084A - Non-criticality measuring device for spent fuel assembly - Google Patents
Non-criticality measuring device for spent fuel assemblyInfo
- Publication number
- JPH0659084A JPH0659084A JP4215038A JP21503892A JPH0659084A JP H0659084 A JPH0659084 A JP H0659084A JP 4215038 A JP4215038 A JP 4215038A JP 21503892 A JP21503892 A JP 21503892A JP H0659084 A JPH0659084 A JP H0659084A
- Authority
- JP
- Japan
- Prior art keywords
- neutron
- spent fuel
- fuel assembly
- count rate
- distribution
- 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.)
- Withdrawn
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Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
(57)【要約】
【目的】 使用済燃料集合体の燃焼度の変動に伴う検出
器検出効率を一定値とし、種々の燃焼度を有する使用済
燃料集合体の未臨界度を精度よく測定可能する。
【構成】 使用済燃料集合体１３をグレイ吸収体１２で
覆われた容器に湿式雰囲気で収容して使用済燃料集合体
１３の水平方向の核分裂中性子分布と内部中性子分布と
をほぼ同一分布にする工夫を施し、装置に近接して第２
の軸方向所定位置中性子計数率を測定する中性子検出器
１５と、第２の軸方向所定位置中性子計数率とあらかじ
め求めてある第１の軸方向所定位置中性子計数率及び集
合体中性子源強度を入力し集合体中性子実効増倍率を演
算する演算手段１６とよりなる。
(57) [Abstract] [Purpose] It is possible to accurately measure the subcriticality of spent fuel assemblies with various burnups, while keeping the detector detection efficiency due to fluctuations in the burnup of spent fuel assemblies to a constant value. To do. [Structure] The spent fuel assembly 13 is housed in a container covered with a gray absorber 12 in a wet atmosphere so that the horizontal fission neutron distribution and the internal neutron distribution of the spent fuel assembly 13 are substantially the same. The device has been devised so that the second
Of the neutron detector 15 for measuring the predetermined axial position neutron count rate, the second predetermined axial position neutron count rate, and the previously determined first predetermined axial position neutron count rate and aggregate neutron source intensity The calculation means 16 calculates the aggregate neutron effective multiplication factor.
Description
【０００１】[0001]
【産業上の利用分野】本発明は、軽水炉などの使用済燃
料集合体の未臨界度測定に係り、特に輸送又は貯蔵体系
内の使用済燃料集合体の燃焼度の変動に伴う検出器検出
効率の変化をなくし、使用済燃料集合体の集合体中性子
実行倍率を求めるのに好適な使用済燃料集合体の未臨界
度測定装置に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the measurement of subcriticality of spent fuel assemblies such as light water reactors, and more particularly, the detection efficiency of detectors due to fluctuations in burnup of spent fuel assemblies in transportation or storage systems. The present invention relates to a device for measuring subcriticality of a spent fuel assembly, which is suitable for obtaining the neutron effective magnification of the spent fuel assembly by eliminating the change in
【０００２】[0002]
【従来の技術】従来のように使用済燃料集合体を水中に
無垢に置いて測定する場合、または使用済燃料集合体に
中性子吸収体を置いて測定する場合は、後述の数２の検
出器検出効率α1が使用済燃料集合体の燃焼度の変動に
伴い変化するため、測定誤差が大きくなっている。2. Description of the Related Art When a spent fuel assembly is placed in pure water for measurement as in the prior art or when a neutron absorber is placed in the spent fuel assembly for measurement, a detector of the following formula 2 is used. Since the detection efficiency α 1 changes with the change in the burnup of the spent fuel assembly, the measurement error becomes large.
【０００３】[0003]
【発明が解決しようとする課題】従来は使用済燃料集合
体を湿式雰囲気で測定する場合、検出器検出効率α1が
使用済燃料集合体の燃焼度の変動に伴い変化するため、
測定誤差が大きくなる問題点があった。Conventionally, when a spent fuel assembly is measured in a wet atmosphere, the detector detection efficiency α 1 changes as the burnup of the spent fuel assembly changes.
There is a problem that the measurement error becomes large.
【０００４】本発明の目的は、燃焼度の変動に伴う検出
器検出効率の変化をなくし、精度よく集合体中性子実行
倍率を求めることのできる使用済燃料集合体の未臨界度
測定装置を提供することにある。An object of the present invention is to provide a subcriticality measuring device for a spent fuel assembly, which can eliminate a change in detector detection efficiency due to a change in burnup and can accurately determine an assembly neutron effective multiplication factor. Especially.
【０００５】[0005]
【課題を解決するための手段】前記の目的を達成するた
め、本発明に係る使用済燃料集合体の未臨界度測定装置
は、使用済燃料集合体をグレイ吸収体で覆われた容器に
湿式雰囲気で収容して使用済燃料集合体の水平方向の核
分裂中性子分布と内部中性子分布とをほぼ同一分布にす
る工夫を施し、装置に近接して第２の軸方向所定位置中
性子計数率を測定する中性子検出器と、第２の軸方向所
定位置中性子計数率とあらかじめ求めてある第１の軸方
向所定位置中性子計数率及び集合体中性子源強度とを入
力し集合体中性子実効増倍率を演算する演算手段とより
なる構成とする。In order to achieve the above-mentioned object, a subcriticality measuring device for a spent fuel assembly according to the present invention wets the spent fuel assembly in a container covered with a gray absorber. Measured to measure the neutron count rate in the second predetermined axial position in the vicinity of the device by making an arrangement so that the nuclear fission neutron distribution in the horizontal direction and the internal neutron distribution of the spent fuel assembly housed in the atmosphere are almost the same. Calculation for calculating the effective neutron multiplication factor of the aggregate by inputting the neutron detector, the second predetermined axial position neutron count rate, and the previously determined first predetermined axial position neutron count rate and aggregate neutron source intensity It is configured by means.
【０００６】そして使用済燃料集合体をグレイ吸収体で
覆われた容器に湿式雰囲気で収容し、グレイ吸収体の周
辺に少なくとも一つの外部中性子源を設け、それぞれの
外部中性子源を、それぞれの外部中性子源による燃料内
熱中性子束分布と核分裂中性子による燃料内熱中性子束
分布とをほぼ同一分布にする位置に配置し、軸方向所定
位置中性子計数率を測定する中性子検出器と、集合体中
性子実効増倍率を演算する演算手段とを具備した構成で
もよい。Then, the spent fuel assemblies are contained in a container covered with a gray absorber in a wet atmosphere, at least one external neutron source is provided around the gray absorber, and each external neutron source is connected to each external neutron source. A neutron detector for arranging the thermal neutron flux distribution in the fuel due to the neutron source and the thermal neutron flux distribution in the fuel due to the fission neutrons at substantially the same distribution, and measuring the neutron count rate at a predetermined axial position, and an aggregate neutron effective It may be configured to include a calculation unit that calculates a multiplication factor.
【０００７】また使用済燃料集合体の周辺に、熱中性子
吸収量を調整する金属又は樹脂で形成された調整部材を
設けた構成でもよい。Further, a structure may be used in which an adjusting member formed of metal or resin for adjusting the amount of thermal neutron absorption is provided around the spent fuel assembly.
【０００８】さらにグレイ吸収体は、金属又は樹脂で形
成されている構成でもよい。Further, the gray absorber may be made of metal or resin.
【０００９】[0009]
【作用】本発明によれば、グレイ吸収体により、内部中
性子源分布とその中性子により誘発される水平方向の核
分裂中性子分布とが同等になるため、燃焼度の変動に伴
う検出器検出効率がほぼ一定値となり、事前にその検出
器検出効率を未照射燃料などについて求めて置くことに
より、容易に第２の軸方向所定位置中性子計数率が測定
され、第１及び第２の軸方向所定位置中性子計数率と、
集合体中性子源強度とに基づいて集合体中性子実効増倍
率（未臨界度）が精度よく演算される。According to the present invention, the gray absorber makes the internal neutron source distribution equal to the horizontal fission neutron distribution induced by the neutrons, so that the detector detection efficiency due to the burnup fluctuation is almost equal. It becomes a constant value, and the detector detection efficiency is obtained in advance for unirradiated fuel etc., and the neutron count rate at the second predetermined axial position neutrons is easily measured. Counting rate,
The effective neutron multiplication factor (subcriticality) is accurately calculated based on the collective neutron source intensity.
【００１０】[0010]
【実施例】本発明の一実施例を図１を参照しながら説明
する。図１に示すように、未臨界度測定装置１１は、中
性子の弱い吸収能力をもつグレイ吸収体１２と、使用済
燃料集合体１３の第２の軸方向所定位置中性子計数率を
測定する中性子検出器１５と、その測定値及びあらかじ
め例えば中性子源強度測定装置で求めた第１の軸方向所
定位置中性子計数率及び内部中性子源強度軸方向分布を
基に数式に基づいて集合体中性子実効増倍率を演算する
演算手段１６とよりなる構成とする。グレイ吸収体１２
は、中性子の弱い吸収力を有しボロンまたはボロン入り
ＳＵＳなどの金属または樹脂で形成されている。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to FIG. As shown in FIG. 1, the subcriticality measuring device 11 is a neutron detector that measures a neutron count rate at a second predetermined axial position of a gray absorber 12 having a weak neutron absorption capacity and a spent fuel assembly 13. The effective neutron multiplication factor based on a mathematical expression based on the instrument 15, its measured value, and the first predetermined axial direction predetermined position neutron count rate and internal neutron source strength axial distribution obtained in advance by, for example, a neutron source strength measuring device. It is configured to include a computing means 16 for computing. Gray absorber 12
Has a weak absorption power for neutrons and is formed of metal or resin such as boron or SUS containing boron.
【００１１】本実施例の動作を説明する。あらかじめ他
の装置を用い、使用済燃料集合体の所定断面位置で第１
の軸方向所定位置中性子計数率Ｃ0を測定しておく。第
１の軸方向所定位置中性子計数Ｃ0は数１に示すように
検出器位置熱中性子束計算値と比例関係にある。The operation of this embodiment will be described. Use another device in advance to make a first measurement at a predetermined cross-sectional position of the spent fuel assembly.
The neutron count rate C 0 at a predetermined axial position is measured. The first predetermined axial position neutron count C 0 is proportional to the detector position thermal neutron flux calculation value as shown in equation 1.
【００１２】[0012]
【数１】 [Equation 1]
【００１３】他の装置では、使用済燃料集合体の燃焼度
の変動に伴う本体系集合体中性子実効増倍率Ｋ0及び検
出器検出効率α0をほぼ一定値とし、集合体中性子源強
度Ｓ0と第１の軸方向所定位置中性子計数率Ｃ0とを比例
関係としているため、容易に集合体中性子源強度Ｓ0が
数１により求められている。In another apparatus, the main body system neutron effective multiplication factor K 0 and the detector detection efficiency α 0 are set to substantially constant values with the burnup of the spent fuel assembly varying, and the assembly neutron source strength S 0 is set. Since the neutron count rate C 0 and the first predetermined axial position neutron count rate are in a proportional relationship, the aggregate neutron source intensity S 0 is easily obtained by the equation 1.
【００１４】数１で求められている集合体中性子源強度
Ｓ0と第１の軸方向所定位置中性子計数率Ｃ0とを基に、
図１に示す未臨界度測定装置で計測した第２の所定位置
中性子計数率Ｃ1は、数２及数３に示すような集合体中
性子実効増倍率Ｋ1との関係を有する。Based on the aggregate neutron source intensity S 0 and the neutron count rate C 0 at the first predetermined axial position, which are obtained by the equation 1,
The second predetermined position neutron count rate C 1 measured by the subcriticality measuring device shown in FIG. 1 has a relationship with the collective neutron effective multiplication factor K 1 as shown in Formulas 2 and 3.
【００１５】[0015]
【数２】 [Equation 2]
【００１６】[0016]
【数３】 [Equation 3]
【００１７】ここでα1とＡとは比例定数であり、α1及
びＡを使用済燃料集合体の燃焼度の変動にかかわらず一
定値にするのはグレイ吸収体であり、グレイ吸収体は燃
焼燃料から発生する内部中性子源分布と水平方向の核分
裂中性子分布を同等にする材料である。解析により数４
によりＡを算出しておく。Here, α 1 and A are proportional constants, and it is the gray absorber that keeps α 1 and A constant regardless of fluctuations in the burnup of the spent fuel assembly, and the gray absorber is It is a material that makes the distribution of internal neutron sources generated from combustion fuel equal to the distribution of horizontal fission neutrons. Number 4 by analysis
A is calculated by.
【００１８】[0018]
【数４】 [Equation 4]
【００１９】測定値Ｃ0，Ｃ1を数３に代入し、集合体中
性子実効倍率Ｋ1を求めることができる。The aggregate neutron effective magnification K 1 can be obtained by substituting the measured values C 0 and C 1 into the equation 3.
【００２０】本発明の他の実施例を図２を参照しながら
説明する。パッシブ中性子増倍法による場合は、別途に
中性子源強度を測定する必要があり、手間がかかる。ア
クティブ中性子増倍法による場合、その外部中性子分布
の影響が未臨界度測定上大きな誤差発生要因となる。図
２に示すように、使用済燃料集合体２３をグレイ吸収体
２２で覆われた容器に湿式雰囲気で収容し、グレイ吸収
体２２の周辺に少なくとも一つの外部中性子源２７を設
け、それぞれの外部中性子源２７を、それぞれの外部中
性子源２７による燃料内熱中性子束分布と核分裂中性子
による燃料内熱中性子束分布とをほぼ同一分布にする位
置に配置し、軸方向所定位置中性子計数率を測定する中
性子検出器２５と、集合体中性子実効増倍率を演算する
演算手段２６とを具備した構成とする。本実施例の効果
をさらに高めるため、使用済燃料集合体の回りに熱中性
子吸収量を調整する図示しない調整部材を設けた構成で
もよい。本調整部材は金属又は樹脂で形成されるものと
する。Another embodiment of the present invention will be described with reference to FIG. In the case of the passive neutron multiplication method, it is necessary to measure the neutron source intensity separately, which is troublesome. In the case of the active neutron multiplication method, the influence of the external neutron distribution becomes a major error factor in subcriticality measurement. As shown in FIG. 2, the spent fuel assembly 23 is contained in a container covered with the gray absorber 22 in a wet atmosphere, and at least one external neutron source 27 is provided around the gray absorber 22 and each external neutron source 27 is provided. The neutron source 27 is arranged at a position where the thermal neutron flux distribution in the fuel by each external neutron source 27 and the thermal neutron flux distribution in the fuel by the fission neutrons are almost the same distribution, and the neutron count rate at a predetermined axial position is measured. The neutron detector 25 and a calculation means 26 for calculating the effective neutron multiplication factor are provided. In order to further enhance the effect of the present embodiment, a configuration may be provided in which an adjusting member (not shown) that adjusts the thermal neutron absorption amount is provided around the spent fuel assembly. This adjusting member is made of metal or resin.
【００２１】[0021]
【発明の効果】本発明によれば、輸送又は貯蔵体系内の
使用済燃料集合体の燃焼度の変動に伴う検出器検出効率
を一定値とすることができ、未臨界度を精度よく測定す
ることができる。EFFECTS OF THE INVENTION According to the present invention, the detection efficiency of the detector can be set to a constant value due to the fluctuation of the burnup of the spent fuel assembly in the transportation or storage system, and the subcriticality can be accurately measured. be able to.
【図１】本発明の一実施例を示す構成図である。FIG. 1 is a configuration diagram showing an embodiment of the present invention.
【図２】本発明の他の実施例を示す構成図である。FIG. 2 is a configuration diagram showing another embodiment of the present invention.
１１ 未臨界度測定装置 １２ グレイ吸収体 １３ 使用済燃料集合体 １５ 中性子検出器 １６ 演算手段 ２１ 未臨界度測定装置 ２２ グレイ吸収体 ２３ 使用済燃料集合体 ２５ 中性子検出器 ２６ 演算手段 11 Subcriticality Measuring Device 12 Gray Absorber 13 Spent Fuel Assembly 15 Neutron Detector 16 Computational Means 21 Subcriticality Measuring Device 22 Gray Absorber 23 Spent Fuel Assembly 25 Neutron Detector 26 Computational Means
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【手続補正書】[Procedure amendment]
【提出日】平成４年８月１８日[Submission date] August 18, 1992
【手続補正２】[Procedure Amendment 2]
【補正対象書類名】明細書[Document name to be amended] Statement
【補正対象項目名】全文[Correction target item name] Full text
【補正方法】変更[Correction method] Change
【補正内容】[Correction content]
【書類名】 明細書[Document name] Statement
【発明の名称】 使用済燃料集合体の未臨界度測定装置Title: Subcriticality measuring device for spent fuel assemblies
【特許請求の範囲】[Claims]
【発明の詳細な説明】Detailed Description of the Invention
【０００１】[0001]
【産業上の利用分野】本発明は、軽水炉などの使用済燃
料集合体の未臨界度測定に係り、特に輸送又は貯蔵体系
内の使用済燃料集合体の燃焼度の変動に伴う検出器検出
効率の変化をなくし、使用済燃料集合体の集合体中性子
実行倍率を求めるのに好適な使用済燃料集合体の未臨界
度測定装置に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the measurement of subcriticality of spent fuel assemblies such as light water reactors, and more particularly, the detection efficiency of detectors due to fluctuations in burnup of spent fuel assemblies in transportation or storage systems. The present invention relates to a device for measuring subcriticality of a spent fuel assembly, which is suitable for obtaining the neutron effective magnification of the spent fuel assembly by eliminating the change in
【０００２】[0002]
【従来の技術】従来のように使用済燃料集合体を水中に
無垢に置いて測定する場合、後述の数２の検出器検出効
率α1が使用済燃料集合体の燃焼度の変動に伴い変化す
るため、測定誤差が大きくなっている。BACKGROUND ART When measuring solid puts the conventional manner spent fuel assemblies in water, detector detection efficiency alpha 1 of 2 number of discussed later is due to the variation of the combustion of the spent fuel assemblies Due to the change, the measurement error is large.
【０００３】[0003]
【発明が解決しようとする課題】従来は使用済燃料集合
体を湿式雰囲気で測定する場合、検出器検出効率α1が
使用済燃料集合体の燃焼度の変動に伴い変化するため、
測定誤差が大きくなる問題点があった。Conventionally, when a spent fuel assembly is measured in a wet atmosphere, the detector detection efficiency α 1 changes as the burnup of the spent fuel assembly changes.
There is a problem that the measurement error becomes large.
【０００４】本発明の目的は、燃焼度の変動に伴う検出
器検出効率の変化をなくし、精度よく集合体中性子実行
倍率を求めることのできる使用済燃料集合体の未臨界度
測定装置を提供することにある。An object of the present invention is to provide a subcriticality measuring device for a spent fuel assembly, which can eliminate a change in detector detection efficiency due to a change in burnup and can accurately determine an assembly neutron effective multiplication factor. Especially.
【０００５】[0005]
【課題を解決するための手段】前記の目的を達成するた
め、本発明に係る使用済燃料集合体の未臨界度測定装置
は、使用済燃料集合体の水平方向の核分裂中性子発生分
布と内部中性子発生分布とをほぼ同一分布にするような
熱中性子吸収能力を有する金属又は樹脂で形成された調
整部材を使用済燃料集合体の周辺に設置し、装置に近接
して第２の軸方向所定位置中性子計数率を測定する中性
子検出器と、第２の軸方向所定位置中性子計数率とあら
かじめ求めてある集合体中性子源強度に係る第１の軸方
向所定位置中性子計数率とを入力し集合体中性子実効増
倍率を演算する演算手段とよりなる構成とする。Means for Solving the Problems] To achieve the above object, subcriticality measuring device of spent fuel assemblies according to the present invention, horizontal fission neutron generator worth of use for spent fuel assemblies <br /> such as to almost the same distribution and cloth and internal neutron generator distribution
A key formed of metal or resin that has the ability to absorb thermal neutrons.
A neutron detector that installs an alignment member around the spent fuel assembly and measures the neutron count rate at the second predetermined axial position in proximity to the device, and obtains the neutron count rate at the second predetermined axial position in advance and it becomes more configuration as the first axial position to enter the neutron count rate calculating means for calculating an aggregate effective neutron multiplication factor of the aggregate neutron source strength in Te.
【０００６】そして外部中性子源による燃料内水平方向
熱中性子束分布と集合体内核分裂中性子による燃料内水
平方向熱中性子束分布とをほぼ同一分布にするような熱
中性子吸収能力を有する金属又は樹脂で形成された調整
部材を使用済燃料集合体の周辺に設置し、軸方向所定位
置中性子計数率を測定する中性子検出器と、集合体中性
子実効増倍率を演算する演算手段とを具備した構成でも
よい。[0006] And the horizontal direction in the fuel by the external neutron source
Thermal neutron flux distribution and water in fuel due to fission neutrons in the assembly
Heat that makes the distribution of the thermal neutron flux in the horizontal direction almost the same.
Conditioning made of metal or resin with neutron absorption capacity
A member may be installed around the spent fuel assembly , and a neutron detector for measuring the neutron count rate at a predetermined axial position and a calculation means for calculating the effective neutron multiplication factor of the assembly may be provided.
【０００７】また調整部材は、グレイ吸収体である構成
でもよい。The adjusting member may be a gray absorber .
【０００８】[0008]
【作用】本発明によれば、グレイ吸収体により、内部中
性子源分布とその中性子により誘発される水平方向の核
分裂中性子分布とが同等になるため、燃焼度の変動に伴
う検出器検出効率がほぼ一定値となり、事前にその検出
器検出効率を未照射燃料などについて求めておくことに
より、第１及び第２の軸方向所定位置中性子計数率に基
づいて容易に集合体中性子実効増倍率（未臨界度）が精
度よく演算される。According to the present invention, the gray absorber makes the internal neutron source distribution equal to the horizontal fission neutron distribution induced by the neutrons, so that the detector detection efficiency due to the burnup fluctuation is almost equal. It becomes a constant value in advance by the detector detection efficiency of fresh fuel, such as for determined Contact Kukoto easily aggregate neutrons group <br/> Zui the first and second axial position neutron count rate The effective multiplication factor (subcriticality) is accurately calculated.
【０００９】[0009]
【実施例】本発明の一実施例を図１を参照しながら説明
する。図１に示すように、未臨界度測定装置１１は、中
性子の弱い吸収能力をもつグレイ吸収体１２と、使用済
燃料集合体１３の第２の軸方向所定位置中性子計数率を
測定する中性子検出器１５と、その測定値及びあらかじ
め例えば中性子源強度測定装置で求めた第１の軸方向所
定位置中性子計数率及び内部中性子源強度軸方向分布を
基に数式に基づいて集合体中性子実効増倍率を演算する
演算手段１６とよりなる構成とする。グレイ吸収体１２
は、例えば、中性子の弱い吸収能力を有するボロン入り
ステンレス鋼などの金属または樹脂で形成されている。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to FIG. As shown in FIG. 1, the subcriticality measuring device 11 is a neutron detector that measures a neutron count rate at a second predetermined axial position of a gray absorber 12 having a weak neutron absorption capacity and a spent fuel assembly 13. The effective neutron multiplication factor based on a mathematical expression based on the instrument 15, its measured value, and the first predetermined axial direction predetermined position neutron count rate and internal neutron source strength axial distribution obtained in advance by, for example, a neutron source strength measuring device. It is configured to include a computing means 16 for computing. Gray absorber 12
Is, for example, a weak absorption capacity of the neutron Yes surbo Ron containing
It is made of metal such as stainless steel or resin.
【００１０】本実施例の動作を説明する。あらかじめ他
の装置を用い、使用済燃料集合体の所定断面位置で第１
の軸方向所定位置中性子計数率Ｃ0を測定しておく。第
１の軸方向所定位置中性子計数Ｃ0は数１に示すように
検出器位置熱中性子束計算値と比例関係にある。The operation of this embodiment will be described. Use another device in advance to make a first measurement at a predetermined cross-sectional position of the spent fuel assembly.
The neutron count rate C 0 at a predetermined axial position is measured. The first predetermined axial position neutron count C 0 is proportional to the detector position thermal neutron flux calculation value as shown in equation 1.
【００１１】[0011]
【数１】 [Equation 1]
【００１２】他の装置では、使用済燃料集合体の燃焼度
の変動に伴う本体系集合体中性子実効増倍率Ｋ0及び検
出器検出効率α0をほぼ一定値とし、集合体中性子源強
度Ｓ0と第１の軸方向所定位置中性子計数率Ｃ0とを比例
関係としているため、容易に集合体中性子源強度Ｓ0が
数１により求められている。In another device, the main body assembly neutron effective multiplication factor K 0 and the detector detection efficiency α 0 are set to substantially constant values with the burnup of the spent fuel assembly changed, and the assembly neutron source strength S 0 is set. Since the neutron count rate C 0 and the first predetermined axial position neutron count rate are in a proportional relationship, the aggregate neutron source intensity S 0 is easily obtained by the equation 1.
【００１３】数１で求められている集合体中性子源強度
Ｓ0と第１の軸方向所定位置中性子計数率Ｃ0とを基に、
図１に示す未臨界度測定装置で計測した第２の所定位置
中性子計数率Ｃ1は、数２及数３に示すような集合体中
性子実効増倍率Ｋ1との関係を有する。Based on the aggregate neutron source intensity S 0 and the first predetermined axial position neutron count rate C 0 obtained by the equation 1,
The second predetermined position neutron count rate C 1 measured by the subcriticality measuring device shown in FIG. 1 has a relationship with the collective neutron effective multiplication factor K 1 as shown in Formulas 2 and 3.
【００１４】[0014]
【数２】 [Equation 2]
【００１５】[0015]
【数３】 [Equation 3]
【００１６】ここでα1とＡとは比例定数であり、α1及
びＡを使用済燃料集合体の燃焼度の変動にかかわらず一
定値にするのはグレイ吸収体であり、グレイ吸収体は燃
焼燃料から発生する内部中性子源分布と水平方向の核分
裂中性子分布を同等にする材料である。解析により数４
によりＡを算出しておく。Here, α 1 and A are proportional constants, and it is the gray absorber that keeps α 1 and A constant regardless of fluctuations in the burnup of the spent fuel assembly, and the gray absorber is It is a material that makes the distribution of internal neutron sources generated from combustion fuel equal to the distribution of horizontal fission neutrons. Number 4 by analysis
A is calculated by.
【００１７】[0017]
【数４】 [Equation 4]
【００１８】測定値Ｃ0，Ｃ1を数３に代入し、集合体中
性子実効倍率Ｋ1を求めることができる。The aggregate neutron effective magnification K 1 can be obtained by substituting the measured values C 0 and C 1 into the equation 3.
【００１９】本発明の他の実施例を図２を参照しながら
説明する。パッシブ中性子増倍法による場合は、別途に
中性子源強度を測定する必要があり、手間がかかる。ア
クティブ中性子増倍法による場合、その外部中性子分布
の影響が未臨界度測定上大きな誤差発生要因となる。図
２に示すように、使用済燃料集合体２３をグレイ吸収体
２２で覆われた容器に湿式雰囲気で収容し、グレイ吸収
体２２の周辺に少なくとも一つの外部中性子源２７を設
け、それぞれの外部中性子源２７を、それぞれの外部中
性子源２７による燃料内熱中性子束分布と核分裂中性子
による燃料内熱中性子束分布とをほぼ同一分布にする位
置に配置し、軸方向所定位置中性子計数率を測定する中
性子検出器２５と、集合体中性子実効増倍率を演算する
演算手段２６とを具備した構成とする。 Another embodiment of the present invention will be described with reference to FIG. In the case of the passive neutron multiplication method, it is necessary to measure the neutron source intensity separately, which is troublesome. In the case of the active neutron multiplication method, the influence of the external neutron distribution becomes a major error factor in subcriticality measurement. As shown in FIG. 2, the spent fuel assembly 23 is contained in a container covered with the gray absorber 22 in a wet atmosphere, and at least one external neutron source 27 is provided around the gray absorber 22 and each external neutron source 27 is provided. The neutron source 27 is arranged at a position where the thermal neutron flux distribution in the fuel by each external neutron source 27 and the thermal neutron flux distribution in the fuel by fission neutrons are made to have substantially the same distribution, and the neutron count rate at a predetermined axial position is measured. a neutron detector 25, shall be the configuration provided with the calculating means 26 for calculating the aggregate effective neutron multiplication factor.
【００２０】[0020]
【発明の効果】本発明によれば、輸送又は貯蔵体系内の
使用済燃料集合体の燃焼度の変動に伴う検出器検出効率
を一定値とすることができ、未臨界度を精度よく測定す
ることができる。EFFECTS OF THE INVENTION According to the present invention, the detection efficiency of the detector can be set to a constant value due to the fluctuation of the burnup of the spent fuel assembly in the transportation or storage system, and the subcriticality can be accurately measured. be able to.
【図面の簡単な説明】[Brief description of drawings]
【図１】本発明の一実施例を示す構成図である。FIG. 1 is a configuration diagram showing an embodiment of the present invention.
【図２】本発明の他の実施例を示す構成図である。FIG. 2 is a configuration diagram showing another embodiment of the present invention.
【符号の説明】 １１ 未臨界度測定装置 １２ グレイ吸収体 １３ 使用済燃料集合体 １５ 中性子検出器 １６ 演算手段 ２１ 未臨界度測定装置 ２２ グレイ吸収体 ２３ 使用済燃料集合体 ２５ 中性子検出器 ２６ 演算手段[Explanation of Codes] 11 Subcriticality Measuring Device 12 Gray Absorber 13 Spent Fuel Assembly 15 Neutron Detector 16 Computational Means 21 Subcriticality Measuring Device 22 Gray Absorber 23 Spent Fuel Assembly 25 Neutron Detector 26 Computing means
Claims (4)
れた容器に湿式雰囲気で収容して前記使用済燃料集合体
の水平方向の核分裂中性子分布と内部中性子分布とをほ
ぼ同一分布にする工夫を施し、装置に近接して第２の軸
方向所定位置中性子計数率を測定する中性子検出器と、
該第２の軸方向所定位置中性子計数率とあらかじめ求め
てある第１の軸方向所定位置中性子計数率及び集合体中
性子源強度とを入力し集合体中性子実効増倍率を演算す
る演算手段とよりなることを特徴とする使用済燃料集合
体の未臨界度測定装置。1. A spent fuel assembly is contained in a container covered with a gray absorber in a wet atmosphere so that the horizontal fission neutron distribution and internal neutron distribution of the spent fuel assembly are substantially the same. A neutron detector, which is devised so as to measure the neutron count rate at a second predetermined axial position in proximity to the device,
The second axial predetermined position neutron count rate and the first predetermined axial position predetermined neutron count rate and the aggregate neutron source intensity, which have been obtained in advance, are included in the calculation means for calculating the effective neutron multiplication factor of the aggregate neutrons. A subcriticality measuring device for a spent fuel assembly, which is characterized in that
れた容器に湿式雰囲気で収容し、前記グレイ吸収体の周
辺に少なくとも一つの外部中性子源を設け、それぞれの
外部中性子源を、それぞれの外部中性子源による燃料内
熱中性子束分布と核分裂中性子による燃料内熱中性子束
分布とをほぼ同一分布にする位置に配置し、軸方向所定
位置中性子計数率を測定する中性子検出器と、集合体中
性子実効増倍率を演算する演算手段とを具備したことを
特徴とする使用済燃料集合体の未臨界度測定装置。2. A spent fuel assembly is housed in a container covered with a gray absorber in a wet atmosphere, and at least one external neutron source is provided around the gray absorber, and each external neutron source is respectively provided. Neutron detector for measuring the neutron count rate at a predetermined axial position by arranging the thermal neutron flux distribution in the fuel due to the external neutron source and the thermal neutron flux distribution in the fuel due to the fission neutrons at a position where the distribution is almost the same. An apparatus for measuring subcriticality of a spent fuel assembly, comprising: a calculation unit that calculates an effective neutron multiplication factor.
収量を調整する金属又は樹脂で形成された調整部材を設
けたことを特徴とする請求項２記載の使用済燃料集合体
の未臨界度測定装置。3. The spent fuel assembly according to claim 2, wherein an adjusting member made of a metal or a resin for adjusting a thermal neutron absorption amount is provided around the spent fuel assembly. Criticality measuring device.
れていることを特徴とする請求項１〜３のいずれか１項
記載の使用済燃料集合体の未臨界測定装置。4. The subcritical measuring device for a spent fuel assembly according to claim 1, wherein the gray absorber is made of metal or resin.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4215038A JPH0659084A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring device for spent fuel assembly |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4215038A JPH0659084A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring device for spent fuel assembly |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH0659084A true JPH0659084A (en) | 1994-03-04 |
Family
ID=16665731
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP4215038A Withdrawn JPH0659084A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring device for spent fuel assembly |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0659084A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN116884657A (en) * | 2023-06-05 | 2023-10-13 | 兰州大学 | A "U" type neutron detection device for spent fuel burnup depth detection |
-
1992
- 1992-08-12 JP JP4215038A patent/JPH0659084A/en not_active Withdrawn
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN116884657A (en) * | 2023-06-05 | 2023-10-13 | 兰州大学 | A "U" type neutron detection device for spent fuel burnup depth detection |
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