JPS6316298A - Nondestructive measuring method of spent nuclear fuel aggregate - Google Patents
Nondestructive measuring method of spent nuclear fuel aggregateInfo
- Publication number
- JPS6316298A JPS6316298A JP61159952A JP15995286A JPS6316298A JP S6316298 A JPS6316298 A JP S6316298A JP 61159952 A JP61159952 A JP 61159952A JP 15995286 A JP15995286 A JP 15995286A JP S6316298 A JPS6316298 A JP S6316298A
- Authority
- JP
- Japan
- Prior art keywords
- fuel assembly
- spent fuel
- neutron
- burnup
- spontaneous
- 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.)
- Pending
Links
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims description 50
- 238000000034 method Methods 0.000 title description 17
- 239000000446 fuel Substances 0.000 claims description 54
- 230000002269 spontaneous effect Effects 0.000 claims description 43
- 238000005259 measurement Methods 0.000 claims description 22
- 238000000691 measurement method Methods 0.000 claims description 21
- 229910052778 Plutonium Inorganic materials 0.000 claims description 19
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims description 19
- 229910052770 Uranium Inorganic materials 0.000 claims description 12
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims description 12
- 238000001816 cooling Methods 0.000 claims description 11
- 238000004458 analytical method Methods 0.000 claims description 10
- 238000000084 gamma-ray spectrum Methods 0.000 claims description 10
- 238000001228 spectrum Methods 0.000 claims description 3
- 239000000463 material Substances 0.000 claims 1
- 238000000429 assembly Methods 0.000 description 23
- 230000000712 assembly Effects 0.000 description 23
- 230000001066 destructive effect Effects 0.000 description 9
- 238000011088 calibration curve Methods 0.000 description 8
- 238000004364 calculation method Methods 0.000 description 7
- 238000001730 gamma-ray spectroscopy Methods 0.000 description 7
- 230000004907 flux Effects 0.000 description 5
- 238000002485 combustion reaction Methods 0.000 description 4
- 230000000694 effects Effects 0.000 description 3
- OYEHPCDNVJXUIW-FTXFMUIASA-N 239Pu Chemical compound [239Pu] OYEHPCDNVJXUIW-FTXFMUIASA-N 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 230000007774 longterm Effects 0.000 description 2
- OYEHPCDNVJXUIW-OIOBTWANSA-N plutonium-241 Chemical compound [241Pu] OYEHPCDNVJXUIW-OIOBTWANSA-N 0.000 description 2
- 238000012958 reprocessing Methods 0.000 description 2
- 230000005262 alpha decay Effects 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 238000009835 boiling Methods 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 238000007796 conventional method Methods 0.000 description 1
- 238000012937 correction Methods 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 230000005251 gamma ray Effects 0.000 description 1
- 238000012544 monitoring process Methods 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 230000035699 permeability Effects 0.000 description 1
- 239000002574 poison Substances 0.000 description 1
- 231100000614 poison Toxicity 0.000 description 1
- 230000000607 poisoning effect Effects 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000001568 sexual effect Effects 0.000 description 1
- 238000010561 standard procedure Methods 0.000 description 1
- JFALSRSLKYAFGM-OIOBTWANSA-N uranium-235 Chemical compound [235U] JFALSRSLKYAFGM-OIOBTWANSA-N 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
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
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
［発明の目的］
（産業上の利用分野〉
本発明は非破壊測定により使用済燃料集合体の核特性の
評価を行なう使用済燃料集合体の非破壊測定方法に関す
る。DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Industrial Application Field) The present invention relates to a method for non-destructive measurement of spent fuel assemblies for evaluating the nuclear properties of spent fuel assemblies by non-destructive measurement.
（従来の技術）
一般に、原子炉から取出された使用済燃料集合体は、一
定期間燃料貯蔵プールに貯蔵され半減期の短い放射能を
減衰させたのち、輸送容器に収納され再処理工場ま、た
は長期貯蔵施設に運ばれる。(Prior art) Generally, spent fuel assemblies taken out from nuclear reactors are stored in fuel storage pools for a certain period of time to attenuate radioactivity with a short half-life, and then stored in transport containers and transported to reprocessing plants. or transported to long-term storage facilities.
このような使用済燃料集合体が原子力発電所から１看出
される時には、切期濃縮度、燃焼度等のデータが再処理
工場あるいは長期貯蔵施設等の使用済燃料集合体の受入
れ側へ渡される。When such a spent fuel assembly is found at a nuclear power plant, data such as cut-off enrichment and burnup are passed to the receiving side of the spent fuel assembly, such as a reprocessing plant or long-term storage facility. .
原子力発電所等の使用済燃料集合体の発送者側は、極力
誤りのないように搬出作業を行なうが、使用済燃料集合
体の受は入れ側では貯蔵や処理の諸工程を通じて確実に
未臨界性を保つため、使用済燃料集合体の燃焼度、核分
裂性核種濃度などについて独自の測定を行なうなどして
誤りがないことを再確認しなければならない。The sender of spent fuel assemblies from nuclear power plants, etc. carries out the transport work with as little error as possible, but when receiving spent fuel assemblies, the receiving side ensures that they are subcritical through various storage and processing steps. In order to maintain compliance, it is necessary to reconfirm that there are no errors by conducting independent measurements of the burnup, fissile nuclide concentration, etc. of spent fuel assemblies.
このような使用済燃料集合体の核特性を評価する方法と
して、従来いわゆるフランス方式（Ｈ，ＤＡＲＲＯＵＺ
ＥＴ、 ｅｔ、ａｌ、 ＩＡＥＡ−２６０／２０１９８
２ ）と***方式％式％
ＡＲＤＡ−２，Ｐ３９６１９８０）の２つの方法が知ら
れている。Conventionally, the so-called French method (H, DARROUZ) has been used as a method to evaluate the nuclear properties of such spent fuel assemblies.
ET, et, al, IAEA-260/20198
Two methods are known: 2) and West German method % formula % ARDA-2, P3961980).
（発明が解決しようとする問題点）
しかしながらこのような従来方法においては、相異なる
方法で燃焼度やプルトニウム濃度等の重要な燃焼パラメ
ータの重複測定ができず、測定値の信頼性に問題があっ
た。(Problems to be solved by the invention) However, in such conventional methods, it is not possible to repeatedly measure important combustion parameters such as burnup and plutonium concentration using different methods, and there is a problem with the reliability of measured values. Ta.
また***方式では中性子増倍率測定法を用いるため、中
性子増倍率の測定のとぎにＣｆ２５２等の外部中性子源
が必要となり、この外部中性子源は半減期の関係から一
定期間ごとに更新しなければならないという問題もあっ
た。Furthermore, since the West German method uses a neutron multiplication rate measurement method, an external neutron source such as Cf252 is required to measure the neutron multiplication rate, and this external neutron source must be updated at regular intervals due to its half-life. There was also the problem.
そこで本発明は、このような問題点８解決するためにな
された方法であり、相異なる方法で燃焼度やプルトニウ
ム濃度等の重要な燃焼パラメータの重複測定をすること
により測定値の信頼性を大幅に向上させ、しかも外部中
性子源を使用しないで中性子増倍率を測定して核***性
核種濃度を求めることが可能な使用済燃料集合体の非破
壊測定方法を提供することを目的とする。Therefore, the present invention is a method developed to solve these problems, and it greatly increases the reliability of measured values by repeatedly measuring important combustion parameters such as burnup and plutonium concentration using different methods. It is an object of the present invention to provide a non-destructive measurement method for spent fuel assemblies that can improve the performance of spent fuel assemblies and determine the concentration of fissile nuclides by measuring the neutron multiplication factor without using an external neutron source.
［発明の構成］
（問題を解決するための手段）
すなわち本発明は、使用済燃料集合体から取出されるガ
ンマ線のスペクトルを測定して分析するガンマ線スペク
トル分析段階と、使用済燃料集合体から放出される自発
中性子の放出率を測定する自発中性子放出率測定段階と
、使用済燃料集合体の外部または内部の少なくとも一方
の中性子減速媒質の少なくとも一部を核的性質の興なる
物質に置き換えることにより使用済燃料集合体から放出
された自発中陣子が増倍される増倍特性を変化させて増
倍率を測定する自発中性子増倍率測定段階とを含み、（
イ）ガンマ線スペクトル分析段階により燃焼度とプルト
ニウム濃度と冷却時間を求めさらに使用済燃料集合体が
ウラン燃料を用いたしのかプルトニウム富化燃料を用い
たものかを判定し、（ロ）使用済燃料集合体がウラン燃
料を用いたものでかつ燃焼度が定められた一定値以上の
ものと使用済燃料集合体がプルトニウム富化燃料を用い
たものが、予め定められた一定値以上の冷却時間を有す
るときには自発中性子放出率測定段階により燃焼度とプ
ルトニウム濃度を導出して使用済燃料集合体の初期濃縮
度を求め、（ハ）一方便用済燃料集合体がプルトニウム
富化燃料を用いたものであるとき、または使用済燃料集
合体がウラン燃料を用いたものでかつガンマ線スペクト
ルの分析段階で求められた燃焼度が予め定められた一定
値以上のときに自発中性子増倍率測定段階により中性子
増倍率を求め、この中性子増倍率から核***性核種濃度
と無限増倍率を導出して使用済燃料集合体の初期濃縮度
を求めるものである。[Structure of the Invention] (Means for Solving the Problem) That is, the present invention includes a gamma ray spectrum analysis step for measuring and analyzing the spectrum of gamma rays extracted from a spent fuel assembly, and a gamma ray spectrum analysis step for measuring and analyzing the spectrum of gamma rays extracted from a spent fuel assembly. a spontaneous neutron emission rate measurement step of measuring the rate of spontaneous neutron emission caused by the spent fuel assembly; a spontaneous neutron multiplication rate measuring step of measuring the multiplication rate by changing the multiplication characteristic by which the spontaneous neutrons released from the spent fuel assembly are multiplied;
b) Determine the burnup, plutonium concentration, and cooling time through the gamma ray spectrum analysis step, and determine whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel, and (b) spend fuel assembly. The spent fuel assembly uses uranium fuel and the burnup is above a predetermined value, and the spent fuel assembly uses plutonium-enriched fuel, and the cooling time is longer than a predetermined value. Sometimes, the initial enrichment of the spent fuel assembly is determined by deriving the burnup and plutonium concentration through a spontaneous neutron emission rate measurement step; (c) On the other hand, the spent fuel assembly uses plutonium enriched fuel. or when the spent fuel assembly uses uranium fuel and the burnup determined in the gamma ray spectrum analysis step is above a predetermined value, the neutron multiplication factor is measured in the spontaneous neutron multiplication factor measurement step. The fissile nuclide concentration and infinite multiplication factor are derived from this neutron multiplication factor to obtain the initial enrichment of the spent fuel assembly.
（作 用）
本発明の使用済燃料集合体の非破壊測定方法は、ガンマ
線スペクトル分析法、自発中性子増倍率測定法、自発中
性子増倍率測定法とを組合せてなり、まずガンマ線スペ
クトル分析法により、燃焼度（ＢｔＪ）とプルトニウム
８度（Ｐｕ／Ｕ）と冷却時間（Ｔｃ）を求め、さらに使
用済燃料集合体がウラン燃料によるものかプルトニウム
富化燃料によるものかを判定し、燃焼度（Ｂｕ）と冷却
時間（Ｔｃ＞を評価しながら自発中性子放出率測定法に
より中性子放出率（Ｓ）と燃焼度（ＢＵ）とプルトニウ
ム濃度（Ｐｕ／Ｕ）を求め、一方自発中性子増倍率測定
法により核***性核種濃度（Ｆｉｓｓ）と実効増倍率（
ｋｅｆｆ）と無限増倍率（ｋＯＯ）を求める。(Function) The method for non-destructive measurement of spent fuel assemblies of the present invention combines gamma ray spectrum analysis, spontaneous neutron multiplication factor measurement, and spontaneous neutron multiplication factor measurement. The burnup (BtJ), plutonium 8 degrees (Pu/U), and cooling time (Tc) are determined, and it is determined whether the spent fuel assembly is made of uranium fuel or plutonium-enriched fuel, and the burnup (Bu/U) is determined. ) and cooling time (Tc>), the neutron emission rate (S), burnup (BU), and plutonium concentration (Pu/U) are determined by the spontaneous neutron emission rate measurement method, while the nuclear fission rate is determined by the spontaneous neutron multiplication rate measurement method. sexual nuclide concentration (Fiss) and effective multiplication factor (
keff) and infinite multiplication factor (kOO).
（実施例） 以下本発明方法の詳細を一実施例について説明する。(Example) The details of the method of the present invention will be explained below with reference to one embodiment.
なお、この実施例では参考データとして使用済燃料集合
体の発送者側から燃料集合体平均の初期濃縮度（εｉ）
、燃焼度（ＢｔＪ）、照射終了年月日等のデータととも
に使用済燃料集合体が受入れ側に送られてきた場合が想
定されている。In this example, as reference data, the average initial enrichment of the fuel assemblies (εi) was obtained from the shipper of the spent fuel assemblies.
It is assumed that a spent fuel assembly is sent to the receiving side along with data such as burn-up (BtJ), date of completion of irradiation, etc.
第１図は本発明の使用済燃料集合体の非破壊測定方法の
一実施例を示すフローチャートである。FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel assemblies of the present invention.
図に示すように、この実施例では、まずガンマ線スペク
トル分析法により燃焼度（ＢＵＤ、プルトニウム濃度（
Ｐｕ／Ｌ濃度比で表すことが多い）、冷却時間（Ｔｃ）
を求め、使用済燃料集合体がウラン燃料を用いたものか
プルトニ・クム富化燃料を用いたものかの判定を行なう
。このガンマ線スペクトル分析法では例えば第２図ない
し第４図に示すように、ホトピーク計数率比Ｃ５１３４
／Ｃｓ１３７、Ｐｒ１４４／Ｃｓ１３７を用いた校正曲
線が用いられる。As shown in the figure, in this example, burnup (BUD) and plutonium concentration (
(often expressed as Pu/L concentration ratio), cooling time (Tc)
, and determine whether the spent fuel assembly uses uranium fuel or Plutoni-Kum enriched fuel. In this gamma ray spectrum analysis method, for example, as shown in FIGS. 2 to 4, the photopeak count rate ratio C5134
A calibration curve using /Cs137 and Pr144/Cs137 is used.
このようにして得られた冷却時間（Ｔｃ）の値が２〜２
．５年以上で、燃焼度（ＢＬＩ）の値が１０〜１５ＧＷ
ｄ／を以上であれば自発中性子放出率測定法を適用する
ことかできる。The value of the cooling time (Tc) obtained in this way is 2 to 2
．． For more than 5 years, the burn-up (BLI) value is 10-15GW
If d/ is greater than or equal to d/, the spontaneous neutron emission rate measurement method can be applied.
また、Ｒｈ１０６から放出されるガンマ線ホトピークの
大きさをモニタすることにより、使用済燃料集合体がウ
ラン燃料を用いたものであるかプルトニウム富化燃料を
用いたものであるかを判断することができる。プルトニ
ウム富化燃料の場合には燃焼度（ＢＵ）の値が低くても
自発中性子放出率測定法を適用することができる。Additionally, by monitoring the size of the gamma ray photopeak emitted from Rh106, it is possible to determine whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel. . In the case of plutonium-enriched fuel, the spontaneous neutron emission rate measurement method can be applied even if the burnup (BU) value is low.
通常の受入れ使用済燃料集合体では、この条件を満足す
るため、発送側から送られてきたデータと照合すること
により大きな誤りのないことを確認することができる。Normally accepted spent fuel assemblies satisfy this condition, so by comparing the data with the data sent from the shipping side, it is possible to confirm that there are no major errors.
自発中性子放出率測定法は、実測値との比較により妥当
性をある程度確認されている五１算コードによりＣｍ２
４２の寄与分を除いた中性子放出率３０４、またはＣｍ
２４４からの中性子放出率Ｓ４と燃焼度（ＢＵＤ、Ｐｕ
全核種合計濃度（ＰＵ）等との相関曲線を、燃料集合体
平均の初期濃縮度（ε１）をパラメータとして第５図に
示すように予め作成しておき、これらが較正曲線として
用いられる。The spontaneous neutron emission rate measurement method uses a 51 calculation code whose validity has been confirmed to some extent by comparison with actual measured values.
Neutron emission rate 304 excluding the contribution of 42, or Cm
Neutron emission rate S4 and burnup (BUD, Pu
A correlation curve with the total concentration of all nuclides (PU), etc. is created in advance as shown in FIG. 5 using the fuel assembly average initial enrichment (ε1) as a parameter, and these curves are used as a calibration curve.
すなわち多数の燃料に対し与えられた燃料集合体平均の
初期濃縮度（εｉ）を用い、自発中性子放出率測定法に
よる中性子放出率から燃焼度（ＢＵ）が求められ、発送
者のデータと比較される。In other words, using the fuel assembly average initial enrichment (εi) given for a large number of fuels, the burnup (BU) is determined from the neutron emission rate by the spontaneous neutron emission rate measurement method, and compared with the sender's data. Ru.
また、１体ずつ燃焼度（ＢＵ）比が求められ、多数の燃
料に対する比の平均値が作成される。そして例えば１５
％以上の著しい差があるものは除外される。この平均値
は予め計算で求められた較正曲線のバイアス値として用
いられ、これにより較正曲線が修正される。Further, the burnup (BU) ratio is determined for each fuel, and an average value of the ratios for a large number of fuels is created. And for example 15
Those with a significant difference of % or more are excluded. This average value is used as a bias value for the calibration curve calculated in advance, and the calibration curve is thereby corrected.
なお、発電所側の燃料の燃焼管理では、１体１体の燃料
集合体の燃焼度（ＢＵ）を正確に求めることは困難であ
るが、多数の燃料集合体合計の出力は電気出力を通して
正確に求められるため、前述のように多数の燃料集合体
に対して求めた比の平均値は非常に信頼性の高いもので
ある。Although it is difficult to accurately determine the burnup (BU) of each fuel assembly in fuel combustion management at the power plant, the total output of many fuel assemblies can be determined accurately through electrical output. Therefore, the average value of the ratio determined for a large number of fuel assemblies as described above is extremely reliable.
このようにして自発中性子放出率測定法の燃焼度（ＢＵ
）に関する修正された較正曲線によりそれぞれの燃料集
合体の燃焼度（ＢｔＪ）が求められる。In this way, the burnup (BU) of the spontaneous neutron emission rate measurement method is
) is used to determine the burnup (BtJ) of each fuel assembly.
またガンマ線スペクトル分析法で得られるホトピーク計
数値は燃焼度（ＢＵ）に比例するが、ガンマ線スペクト
ル分析法によりその比例係数を求めることが困難な場合
がある。このような場合には、ガンマ線スペクトル分析
法で求められたＣ５１３７のホトピーク計数値から燃焼
度＜ＢＵ）を決定するための比例係数を、自発中性子放
出出側定法で得られた燃焼度（ＢＵ）と一致するように
決定することもできる。Further, although the photopeak count value obtained by gamma ray spectroscopy is proportional to the burnup (BU), it may be difficult to obtain the proportionality coefficient by gamma ray spectroscopy. In such a case, the proportional coefficient for determining the burnup <BU) from the photopeak count value of C5137 determined by gamma ray spectroscopy should be replaced by the burnup (BU) obtained by the spontaneous neutron emission standard method. It can also be determined to match.
Ｐｕ全核種合計濃度（Ｐｕ）はガンマ線スペクトル分析
法および自発中性子放出率測定法のいずれの方法によっ
ても得ることができるため、得られた結果を総合比較し
て、より信頼性の高いものとすることができる。The total concentration of all Pu nuclides (Pu) can be obtained by both gamma ray spectroscopy and spontaneous neutron emission rate measurement, so the obtained results should be comprehensively compared to make them more reliable. be able to.
燃料集合体１体１体の燃料集合体平均の初期濃縮度（ε
ｉ）は、ガンマ線スペクトル分析法により求められた燃
焼度（ＢＵ）と自発中性子放出率測定法により求められ
た燃焼度（ＢＵ）とが一致する燃料集合体平均の初期濃
縮度（εｉ）として両者の比較から決定される。燃料集
合体平均の初期濃縮度（εｉ）の種類は通常非常に限ら
れているため、容易に識別決定することができる。The average initial enrichment of each fuel assembly (ε
i) is the fuel assembly average initial enrichment (εi) at which the burnup (BU) determined by gamma ray spectroscopy and the burnup (BU) determined by spontaneous neutron emission rate measurement method match. Determined from a comparison of Since the types of fuel assembly average initial enrichment (εi) are usually very limited, they can be easily identified and determined.
燃料集合体１体１体の核***性核種濃度（ＦｉＳＳ）は
、計算で求めた核***性核種濃度（ＦｉＳＳ）と燃焼度
（ＢＵ）との関係、または核***性核種濃度（Ｆ ｉ
ｓｓ）とＰｕ全核種合計ａ度（Ｐｕ）との相関曲線を用
いて決定することができる。この核***性核種濃度（Ｆ
ｉ ＳＳ）としては、全体′ａ度、ウラン２３５１度
、プルトニウム２３９溌度、プルトニウム２４１濃度等
のいずれでもよい。The fissile nuclide concentration (FiSS) of one fuel assembly is determined by the relationship between the calculated fissile nuclide concentration (FiSS) and burnup (BU), or the fissile nuclide concentration (FiSS) determined by calculation.
It can be determined using a correlation curve between Pu ss) and the total a degree of all Pu nuclides (Pu). This fissile nuclide concentration (F
i SS) may be any of the total 'a degrees, uranium 2351 degrees, plutonium 239 permeability, plutonium 241 concentration, etc.
冷却時間（Ｔｃ＞が約２年以下の場合あるいは燃焼度（
Ｂｔｕ）が１０〜１５ＧＷｄ／を以下（但しプルトニウ
ム富化燃料ではさらに低い燃焼度でもよい）の場合には
、自発中性子放出率測定法の適用は困難となる。If the cooling time (Tc> is about 2 years or less or the burnup (
Btu) is less than 10 to 15 GWd/ (however, even lower burnup is possible for plutonium-enriched fuel), it becomes difficult to apply the spontaneous neutron emission rate measurement method.
一方自発中性子増倍率測定法では、冷却時間（ＴＣ＞の
長さには無関係であり、自発中性子放出率の測定が可能
であればよい。プルトニウム富化燃料では燃焼度が小さ
くてもＰｕ２４０等から放出される中性子の測定が可能
であり、この自発中性子増倍率測定法が実施できる。し
かしウラン燃料の場合には少なくとも５ＧＷｄ／を以上
の燃焼度が達成されていないと自発中性子放出率の測定
が困難となるため、ガンマ線スペクトル分析法の結果し
か得られない。したがってこの場合には重要なパラメー
タの異なる方法により測定ができなくなるので通常は特
別扱いをする。このような燃料集合体としては、例えば
沸騰水型原子炉（ＢＷＲ＞の燃料集合体において可燃性
毒物（Ｇｄ）の中性子毒作用のみが消滅し、Ｕ２３５の
減耗が余り進んでいない破損燃料集合体のようなものが
考えられるが、このような燃料集合体は臨界安全性の観
点からは当然特殊扱いすべきものである。On the other hand, the spontaneous neutron multiplication rate measurement method is unrelated to the length of the cooling time (TC>), as long as it is possible to measure the spontaneous neutron emission rate.For plutonium-enriched fuel, even if the burnup is small, it can be It is possible to measure the emitted neutrons, and this spontaneous neutron multiplication rate measurement method can be implemented.However, in the case of uranium fuel, measurement of the spontaneous neutron emission rate is not possible unless a burnup of at least 5 GWd/ has been achieved. Due to this difficulty, only the results of gamma ray spectroscopy can be obtained.Therefore, in this case, special treatment is usually given as important parameters cannot be measured using different methods.Such fuel assemblies include, for example, In the fuel assembly of a boiling water reactor (BWR), only the neutron poisoning effect of the burnable poison (Gd) has disappeared, and the U235 depletion has not progressed very much in a damaged fuel assembly. Such fuel assemblies should of course be treated specially from the viewpoint of criticality safety.
自発中性子増倍率測定法では使用済燃料集合体内部から
自発俟***によって放出される中性子と、燃料部に含ま
れる酸素と超ウラン核種のアルファ崩壊で放出されるア
ルファ線との（α、ｎ）反応で放出される中性子が中性
子源（自発中性子源、１次中・［４子源、固有の中性子
源、内部中性子源等と呼ばれる）となるのでＣｆ２５２
のような外部中性子源（燃料集合体内部に配置しても燃
料集合体固有の自発中１子源ではないため、外部中性子
源と呼ばれる）を使用する必要がない。外部中性子源と
しては何種類かのものがあるが、その半減期の関係から
いずれも一定期間ごとに更新をしなければならないとい
う問題がおる。たとえば外部中性子源に最適とされてい
るＣｆ２５２中性子源の半減期は２．６５年と比較的短
いので短期間ごとに更新する必要がある。In the spontaneous neutron multiplication rate measurement method, the (α, n) difference between neutrons emitted from inside the spent fuel assembly by spontaneous fission and alpha rays emitted by the alpha decay of oxygen and transuranium nuclides contained in the fuel part is calculated. Since the neutrons released in the reaction become a neutron source (called spontaneous neutron source, primary medium/4 neutron source, intrinsic neutron source, internal neutron source, etc.), Cf252
There is no need to use an external neutron source (called an external neutron source because it is not a spontaneous neutron source specific to the fuel assembly even if it is placed inside the fuel assembly). There are several types of external neutron sources, but all have the problem of having to be updated at regular intervals due to their half-lives. For example, the half-life of the Cf252 neutron source, which is considered optimal as an external neutron source, is relatively short at 2.65 years, so it must be renewed every short period.
自発中性子源の場合には燃焼度の小さいウラン燃料の場
合を除いて中性子源の更新という問題はない。In the case of spontaneous neutron sources, there is no problem of updating the neutron source, except in the case of uranium fuel with a low burnup.
自発中性子を中性子源とする自発中性子増倍率測定法で
は、燃料集合体の内部または外部の少なくとも一方の媒
質の少なくとも１部を燃料以外の媒質に置き換えること
によって燃料集合体の増倍特性を変化させ、それによっ
て変化する中性子束の変化から増倍率を決定する方法で
あり、すでに特開昭５２−６３５９５号公報、特開昭５
３−５２８９３＠公報、ＬＡＮＬ−９４９４−ＭＳ、
ＥＰＲＩ−ＮＰ−２８１８に開示されている。In the spontaneous neutron multiplication factor measurement method using spontaneous neutrons as a neutron source, the multiplication characteristics of the fuel assembly are changed by replacing at least part of the medium inside or outside the fuel assembly with a medium other than fuel. , is a method of determining the multiplication factor from the change in the neutron flux that changes thereby, and has already been disclosed in Japanese Patent Application Laid-Open No. 52-63595 and Japanese Patent Application Laid-open No. 52-63595.
3-52893@publication, LANL-9494-MS,
It is disclosed in EPRI-NP-2818.
自発中性子増倍率測定法により例えば水中に１体だけ使
用済燃料集合体が置かれた場合の実効増倍率（ｋｅｆｆ
）が求められまた核***性核種濃度（Ｆ ｉ ｓｓ）を
求めることもできる。燃料集合体固有の特性を表す無限
増倍率（ｋｏｏ）は、計算で求められる換算係数（中性
子のもれない確率）を用いて無限増倍率（ｋｏｏ）から
算出される。自発中性子増倍率測定法で求められた無限
増倍率（ｋＯｏ）や核***性核種濃度（Ｆｉｓｓ）等の
値は自発中性子放出率測定法において必要な増倍効果の
補正や計算によって求めた較正曲線を用いて半実験的に
導出された前述の無限増倍率（ｋｏｏ）、核***性核種
濃度（Ｆ−ｉｓｓ）などの値の妥当性の評価に使用する
こともできる。Using the spontaneous neutron multiplication factor measurement method, for example, the effective multiplication factor (keff) when only one spent fuel assembly is placed underwater
) can also be determined, and the fissile nuclide concentration (F i ss) can also be determined. The infinite multiplication factor (koo) representing the unique characteristics of the fuel assembly is calculated from the infinite multiplication factor (koo) using a conversion factor (probability that neutrons will not leak) obtained by calculation. Values such as the infinite multiplication factor (kOo) and fissile nuclide concentration (Fiss) obtained by the spontaneous neutron multiplication rate measurement method are based on the correction of the multiplication effect necessary for the spontaneous neutron emission rate measurement method and the calibration curve obtained by calculation. It can also be used to evaluate the validity of values such as the above-mentioned infinite multiplication factor (koo) and fissile nuclide concentration (F-iss) derived semi-experimentally.
自発中性子増倍率測定法では体系の媒質を変化させる前
の中性子束（φ０）と変化させた後の中性子束（φ）と
の比（φ／φ０）が実効増倍率（ｋｅｆｆ）や無限増倍
率（ｋｏｏ） ト第６図のような関係にある特性を利用
する。この相関特性は通常計詐で求められるが、その妥
当性を実験により評価することもできる。相関曲線の作
成にあたっては、組成既知または実効増倍率（ｋｅｆｆ
）既知の燃料集合体を使用する。In the spontaneous neutron multiplication factor measurement method, the ratio (φ/φ0) of the neutron flux (φ0) before changing the system medium to the neutron flux (φ) after changing it is the effective multiplication factor (keff) or infinite multiplication factor. (koo) Utilizes the characteristics having the relationship shown in Figure 6. This correlation characteristic is usually obtained by calculation, but its validity can also be evaluated by experiment. When creating a correlation curve, the composition must be known or the effective multiplication factor (keff
) using known fuel assemblies.
すなわち、まず測定された中性子束比（φ／ψＯ）に基
づいて実効増倍率（ｋｅｆｆ）および核***性核種濃度
（Ｆｉ ｓｓ）が決定される。この実効増倍率（ｋｅｆ
ｆ）に基づく計算により無限増倍率（ｋＯｏ）が決定さ
れる。なお計算により予め中性子束比（φ／φ０）に対
する無限増倍率（ｋｏＯ）の相関関係が燃料集合体平均
の初期濃縮度（εｉ）をパラメータとする較正曲線とし
て、第８図に示すように作成されており、この自発中性
子増倍率測定法で求められた無限増倍率（ｋＣ／Ｑ）
−と、ガンマ線スペクトル分析法で求められた燃
焼度（８ｔＪ）とを用いて燃料集合体平均の初期濃縮度
（εｉ）が決定される。。That is, first, the effective multiplication factor (keff) and the fissile nuclide concentration (Fi ss) are determined based on the measured neutron flux ratio (φ/ψO). This effective multiplication factor (kef
The infinite multiplication factor (kOo) is determined by calculation based on f). By calculation, the correlation between the infinite multiplication factor (koO) and the neutron flux ratio (φ/φ0) was created as a calibration curve with the fuel assembly average initial enrichment (εi) as a parameter, as shown in Figure 8. The infinite multiplication factor (kC/Q) determined by this spontaneous neutron multiplication factor measurement method is
- and the burnup (8 tJ) determined by gamma ray spectroscopy, the fuel assembly average initial enrichment (εi) is determined. .
核***性核種濃度（Ｆｉｓｓ）、すなわちウラン２３５
、プルトニウム２３９およびプルトニウム２４１の合計
濃度は自発中性子増倍率測定法により直接求めることが
できるが、各核種の濃度は予め計算で求めた例えば、第
７図に示すような較正曲線が利用される。また、Ｐｕ全
核種合計）農度（Ｐｕ）も計算で予め求められたＰｕ全
咳種合計濃度（Ｐｕ）と核***性核種濃度（Ｆ ｉ Ｓ
Ｓ）との燃料集合体平均の初ｍｍ縮度（εｉ）をパラメ
ータとした較正曲線より決定される。Fissile nuclide concentration (Fiss), i.e. uranium-235
The total concentration of plutonium 239 and plutonium 241 can be directly determined by spontaneous neutron multiplication factor measurement, but the concentration of each nuclide is determined using a calibration curve as shown in FIG. 7, for example, calculated in advance. In addition, the total concentration of Pu all nuclides (Pu) and the concentration of fissile nuclides (Pu) calculated in advance are calculated as follows:
S) is determined from a calibration curve using the average initial mm shrinkage (εi) of the fuel assembly as a parameter.
以上述べたようにして使用済燃料集合体の非破壊測定が
終了する。この後、各種データを総合し、発送者からの
データと比較することにより、受入れ側の使用済燃料集
合体の管理を安全確実なものとすることができる。As described above, the non-destructive measurement of the spent fuel assembly is completed. Thereafter, by integrating the various data and comparing it with the data from the sender, the spent fuel assembly on the receiving side can be managed safely and reliably.
［発明の効果］
以上述べたように本発明の使用済燃料集合体の非破壊測
定方法によれば、ガンマ線スペクトル分析法と自発中性
子放出率測定法と自発中性子増倍率測定法とを組合ける
ことにより、使用済燃料集合体の冷却時間の大小に関係
なく使用済燃料集合体の非破壊測定を確実に行なうこと
ができ、また燃焼度やプルトニウム濃度等の特に重要な
燃焼パラメータを重複測定することが可能となるので測
定値の信頼性が大幅に向上し、ざらには外部中性子源を
使用せずに中性子増倍率を測定することができるので一
定期間ごとに外部中性子源の更新をする必要がないとい
う効果もある。[Effects of the Invention] As described above, according to the method for non-destructive measurement of spent fuel assemblies of the present invention, gamma ray spectrum analysis, spontaneous neutron emission rate measurement, and spontaneous neutron multiplication rate measurement can be combined. This makes it possible to reliably perform non-destructive measurements of spent fuel assemblies regardless of the cooling time of the spent fuel assemblies, and to double measure especially important combustion parameters such as burnup and plutonium concentration. This greatly improves the reliability of measured values, and in general, it is possible to measure the neutron multiplication factor without using an external neutron source, which eliminates the need to update the external neutron source at regular intervals. There is also the effect of not having one.
第１図は本発明の使用済燃料集合体の非破壊測定方法の
一実施例を示すフローチャート、第２図〜第８図は本発
明で用いられる各種相関曲線を概念的に示すグラフであ
る。
第２図 第３図
第４Ｍ 第５１！
Ｊ
第８図FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel assemblies of the present invention, and FIGS. 2 to 8 are graphs conceptually showing various correlation curves used in the present invention. Figure 2 Figure 3 Figure 4M 51st! J Figure 8
Claims (1)
クトルを測定して分析するガンマ線スペクトル分析段階
と、前記使用済燃料集合体から放出される自発中性子の
放出率を測定する自発中性子放出率測定段階と、前記使
用済燃料集合体の外部または内部の少なくとも一方の中
性子減速媒質の少なくとも一部を核的性質の異なる物質
に置き換えることにより前記使用済燃料集合体から放出
された自発中性子が増倍される増倍特性を変化させて増
倍率を測定する自発中性子増倍率測定段階とを含み、（
イ）前記ガンマ線スペクトル分析段階により燃焼度とプ
ルトニウム濃度と冷却時間を求めさらに前記使用済燃料
集合体がウラン燃料を用いたものかプルトニウム富化燃
料を用いたものかを判定し、（ロ）前記使用済燃料集合
体がウラン燃料を用いたものでかつ燃焼度が定められた
一定値以上のものと前記使用済燃料集合体がプルトニウ
ム富化燃料を用いたものが、予め定められた一定値以上
の前記冷却時間を有するときには前記自発中性子放出率
測定段階により燃焼度とプルトニウム濃度を導出して前
記使用済燃料集合体の初期濃縮度を求め、（ハ）一方前
記使用済燃料集合体がプルトニウム富化燃料を用いたも
のであるとき、または前記使用済燃料集合体がウラン燃
料を用いたものでかつ前記ガンマ線スペクトル分析段階
で求められた前記燃焼度が予め定められた一定値以上の
ときに前記自発中性子増倍率測定段階により中性子増倍
率を求め、この中性子増倍率から核***性核種濃度と無
限増倍率を導出して前記使用済燃料集合体の初期濃縮度
を求めることを特徴とする使用済燃料集合体の非破壊測
定方法。(1) A gamma ray spectrum analysis step of measuring and analyzing the spectrum of gamma rays emitted from the spent fuel assembly, and spontaneous neutron emission rate measurement of measuring the emission rate of spontaneous neutrons emitted from the spent fuel assembly. and multiplying the spontaneous neutrons emitted from the spent fuel assembly by replacing at least a portion of at least one of the neutron moderating medium outside or inside the spent fuel assembly with a material having different nuclear properties. a spontaneous neutron multiplication rate measurement step of measuring the multiplication rate by changing the multiplication characteristics
(b) determining the burnup, plutonium concentration, and cooling time through the gamma ray spectrum analysis step; and determining whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel; The spent fuel assembly uses uranium fuel and the burnup is above a predetermined certain value, and the spent fuel assembly uses plutonium-enriched fuel and the burnup is above a predetermined certain value. (c) when the spent fuel assembly has the cooling time of or when the spent fuel assembly uses uranium fuel and the burnup determined in the gamma ray spectrum analysis step is equal to or higher than a predetermined value. A spent fuel characterized in that a neutron multiplication factor is determined by a spontaneous neutron multiplication factor measurement step, and a fissile nuclide concentration and an infinite multiplication factor are derived from this neutron multiplication factor to determine the initial enrichment of the spent fuel assembly. Non-destructive measurement method for aggregates.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61159952A JPS6316298A (en) | 1986-07-08 | 1986-07-08 | Nondestructive measuring method of spent nuclear fuel aggregate |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61159952A JPS6316298A (en) | 1986-07-08 | 1986-07-08 | Nondestructive measuring method of spent nuclear fuel aggregate |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS6316298A true JPS6316298A (en) | 1988-01-23 |
Family
ID=15704750
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61159952A Pending JPS6316298A (en) | 1986-07-08 | 1986-07-08 | Nondestructive measuring method of spent nuclear fuel aggregate |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6316298A (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2012220297A (en) * | 2011-04-07 | 2012-11-12 | Toshiba Corp | Reactor fuel nondestructive burnup evaluation method and apparatus therefor |
| JP2014185993A (en) * | 2013-03-25 | 2014-10-02 | Toshiba Corp | Evaluation device, evaluation method, and program for burn-up of nuclear fuel |
| JP2015227817A (en) * | 2014-05-30 | 2015-12-17 | 株式会社東芝 | Burnup measurement apparatus of fuel debris and burnup measurement method of the same |
-
1986
- 1986-07-08 JP JP61159952A patent/JPS6316298A/en active Pending
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2012220297A (en) * | 2011-04-07 | 2012-11-12 | Toshiba Corp | Reactor fuel nondestructive burnup evaluation method and apparatus therefor |
| JP2014185993A (en) * | 2013-03-25 | 2014-10-02 | Toshiba Corp | Evaluation device, evaluation method, and program for burn-up of nuclear fuel |
| JP2015227817A (en) * | 2014-05-30 | 2015-12-17 | 株式会社東芝 | Burnup measurement apparatus of fuel debris and burnup measurement method of the same |
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