JPH0659085A - Non-criticality measuring system for spent fuel assembly and non-criticality measurement - Google Patents
Non-criticality measuring system for spent fuel assembly and non-criticality measurementInfo
- Publication number
- JPH0659085A JPH0659085A JP4215039A JP21503992A JPH0659085A JP H0659085 A JPH0659085 A JP H0659085A JP 4215039 A JP4215039 A JP 4215039A JP 21503992 A JP21503992 A JP 21503992A JP H0659085 A JPH0659085 A JP H0659085A
- Authority
- JP
- Japan
- Prior art keywords
- neutron
- spent fuel
- fuel assembly
- count rate
- aggregate
- 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- Withdrawn
Links
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
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
料集合体の燃焼に伴う中性子実効増倍率の低下を考慮す
る臨界管理（燃焼度クレジット）における未臨界度測定
に係り、特に輸送又は貯蔵体系内の使用済燃料集合体の
未臨界度を測定するのに好適な使用済燃料集合体の未臨
界度測定システム及び未臨界度測定方法に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to subcriticality measurement in criticality control (burnup credit) in consideration of reduction in effective neutron multiplication factor associated with combustion of spent fuel assemblies such as light water reactors, and particularly to transportation or The present invention relates to a subcriticality measuring system and a subcriticality measuring method for a spent fuel assembly suitable for measuring the subcriticality of a spent fuel assembly in a storage system.
【０００２】[0002]
【従来の技術】使用済燃料の取扱い施設及び輸送もしく
は貯蔵用容器（キャスク）の臨界安全設計ににおける未
臨界性の評価では、中性子増倍率が最も高くなるような
燃料を想定する必要がある。従来、この想定において未
燃焼の燃料を前提とした臨界安全設計が行われている。2. Description of the Related Art In the evaluation of subcriticality in the criticality safety design of facilities for handling spent fuel and transportation or storage containers (casks), it is necessary to assume a fuel having the highest neutron multiplication factor. Conventionally, in this assumption, a criticality safety design based on unburned fuel has been performed.
【０００３】実際には使用済燃料は、燃焼によって実効
的な核***性物質量が減少することにより核***の確率
が小さくなり、さらに、核***生成物の蓄積により中性
子吸収効果が増大するために、中性子増倍率がが低下し
ている。使用済燃料の取扱い施設の臨界安全設計及び臨
界安全管理において、燃焼に伴う燃料の中性子増倍率の
低下を考慮することを、燃焼度クレジットといい、近
年、燃焼度クレジットの考え方を採用することにつして
の技術的検討が世界各国でなされている。多数の原子力
発電所を有する米国では、既に一部の発電所の燃料貯蔵
プールに燃焼度クレジットを取り入れ、貯蔵容量を増加
させている。In practice, spent fuel has a low probability of nuclear fission due to a decrease in the effective amount of fissile material due to combustion, and the neutron absorption effect increases due to the accumulation of fission products. The multiplication factor is decreasing. In the criticality safety design and criticality safety management of facilities that handle spent fuel, taking into consideration the decrease in the neutron multiplication factor of fuel due to combustion is called burnup credit, and in recent years, the concept of burnup credit has been adopted. As a result, technical studies are being made around the world. In the United States, which has a large number of nuclear power plants, some have already incorporated burnup credits into their fuel storage pools to increase storage capacity.
【０００４】ＡＮＳＩ／ＡＮＳ−５７．７−１９８１
は、使用済燃料貯蔵プールの臨界安全性評価における燃
料の想定を以下の二つの分類のどちらかとしている。ANSI / ANS-57.7-1981
Defines fuel assumptions for criticality safety assessment of spent fuel storage pools as one of the following two categories.
【０００５】ａ）燃焼による核***性物質量の減少は考
慮せず、フール中において中性子増倍率が最も高くなる
ような燃料とすること。A) A fuel that has the highest neutron multiplication factor in the fuel is taken into consideration without considering the decrease in the amount of fissile material due to combustion.
【０００６】ｂ）燃焼による核***性物質量の減少を考
慮する場合には、貯蔵する燃料の最小燃焼度を設定し、
それに対応した中性子増倍率を想定すること、及び実測
により中性子増倍率の低下を確認すること。B) When considering the decrease in the amount of fissile material due to combustion, the minimum burnup of the fuel to be stored is set,
Assuming a neutron multiplication factor corresponding to it, and confirming a decrease in neutron multiplication factor by actual measurement.
【０００７】[0007]
【発明が解決しようとする課題】従来は使用済燃料集合
体の反応度を安全側に高く見積って本体系の臨界解析を
行っているため、臨界安全側へ余裕が大きく、精度に問
題点があった。Conventionally, since the reactivity of the spent fuel assembly is highly estimated on the safe side to perform the criticality analysis of the main body system, there is a large margin on the criticality safe side and there is a problem in accuracy. there were.
【０００８】本発明の目的は、種々の燃焼度を有し、燃
焼度などの運転歴データが不明でも未臨界度の測定を行
うことのできる使用済燃料集合体の未臨界度測定システ
ム及び未臨界度測定方法を提供し、燃焼度クレジットの
導入に寄与ことにある。An object of the present invention is to provide a subcriticality measuring system for a spent fuel assembly which has various burnups and is capable of measuring subcriticality even if operating history data such as burnup is unknown. To provide a method for measuring criticality and contribute to the introduction of burnup credits.
【０００９】[0009]
【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シス
テムは、使用済燃料集合体を収容し該使用済燃料集合体
の燃焼度の変動による係数率の変化を防止する工夫を施
して第１の軸方向所定位置中性子計数率を測定し、第１
の軸方向所定位置中性子計数率に基づき集合体中性子源
強度を演算する中性子源強度測定装置と、使用済燃料集
合体を収容し使用済燃料集合体の水平方向の核***中性
子分布を内部中性子分布とほぼ同一分布にする工夫を施
して第２の軸方向所定位置中性子計数率を測定し、第２
の軸方向所定位置中性子計数率、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に基づき集合体中
性子実効増倍率を演算する未臨界度測定装置とを備えた
構成とする。In order to achieve the above object, a subcriticality measuring system for a spent fuel assembly according to the present invention stores a spent fuel assembly and burns the spent fuel assembly. Measure the neutron count rate at the first predetermined axial position by taking measures to prevent the coefficient rate from changing due to fluctuations in
Neutron source intensity measuring device that calculates the aggregate neutron source intensity based on the axial predetermined position neutron count rate, and the horizontal fission neutron distribution of the spent fuel assembly containing the spent fuel assembly and the internal neutron distribution Measured the neutron count rate at the second predetermined axial position by devising an almost uniform distribution.
And a subcriticality measuring device for calculating the effective neutron multiplication factor of the aggregate neutrons based on the axial predetermined position neutron count rate, the first axial predetermined position neutron count rate, and the aggregate neutron source intensity.
【００１０】そして使用済燃料集合体の未臨界度測定方
法においては、使用済燃料集合体の燃焼度の変動による
係数率の変化を防止する工夫を施して第１の軸方向所定
位置中性子計数率を測定し、第１の軸方向所定位置中性
子計数率に基づき集合体中性子源強度を演算する第１の
手順と、使用済燃料集合体の水平方向の核***中性子分
布を内部中性子分布とほぼ同一分布にする工夫を施して
第２の軸方向所定位置中性子計数率を測定し、第２の軸
方向所定位置中性子計数率、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に基づき集合体中性子
実効増倍率を演算する第２の手順とよりなる構成とす
る。In the method for measuring the subcriticality of a spent fuel assembly, the first axial predetermined position neutron count rate is devised by preventing the change of the coefficient rate due to the variation of the burnup of the spent fuel assembly. And the first procedure for calculating the aggregate neutron source intensity based on the first predetermined axial position neutron count rate, and the horizontal fission neutron distribution of the spent fuel assembly is almost the same as the internal neutron distribution. The second predetermined axial position neutron count rate is measured by devising the following, and the second predetermined axial position neutron count rate, the first predetermined axial position neutron count rate, and the aggregate based on the aggregate neutron source intensity are collected. It is configured to include a second procedure for calculating the effective neutron multiplication factor.
【００１１】[0011]
【作用】本発明によれば、中性子源強度測定装置では、
使用済燃料集合体が空気等のガス雰囲気中で水密内容器
に設置されるため体系内で中性子の熱化が生じないこ
と、かつ容器の外周に中性子吸収体が設けられ、装置が
水中に浸されているため、中性子が使用済燃料集合体の
外周の水部により減速されて容器に入る熱中性子が吸収
される。したがって燃料部の熱中性子がなくなり、本体
系集合体中性子実効増倍率及び検出器検出効率がほぼ一
定値となる。各軸方向の集合体内部中性子発生率（集合
体内部中性子源強度）と第１の軸方向所定位置中性子計
数率とが比例関係を有するようになり、容易に第１の軸
方向所定位置中性子計数率が測定され、測定値より集合
体中性子源強度及び軸方向分布が演算される。According to the present invention, in the neutron source intensity measuring device,
Since the spent fuel assembly is installed in a watertight inner container in a gas atmosphere such as air, neutron thermalization does not occur in the system, and a neutron absorber is installed on the outer periphery of the container so that the device can be immersed in water. Therefore, the neutrons are decelerated by the water portion on the outer periphery of the spent fuel assembly and the thermal neutrons entering the container are absorbed. Therefore, the thermal neutrons in the fuel portion disappear, and the effective neutron multiplication factor of the main body system and the detection efficiency of the detector become substantially constant. The aggregate internal neutron production rate (aggregate internal neutron source intensity) in each axial direction has a proportional relationship with the first axial predetermined position neutron count rate, and the first axial predetermined position neutron count can be easily performed. The rate is measured, and the aggregate neutron source intensity and axial distribution are calculated from the measured values.
【００１２】つぎに未臨界度測定装置では、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***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、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。Next, in the subcriticality measuring device, the gray absorber makes the internal neutron source distribution equal to the horizontal fission neutron distribution induced by the neutrons.
The detector detection efficiency due to the burnup fluctuation becomes almost constant, and the detector detection efficiency of the second axial predetermined position neutron count rate can be easily measured by obtaining the detector detection efficiency for unirradiated fuel in advance. , The first and second axial predetermined position neutron count rates and the collective neutron source intensity are used to accurately calculate the collective neutron effective multiplication factor (subcriticality).
【００１３】[0013]
【実施例】本発明の一実施例を図１及び図２を参照しな
がら説明する。図１に示すように、中性子源強度測定装
置（第１の装置）１は、中性子吸収体２を張付けた水密
容器と、使用済燃料集合体３の第１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５と、その測定値を
基に数式に基づいて集合体中性子源強度を演算する演算
手段６とよりなる構成とする。水密容器は、図示しない
が中性子吸収体２と同一の場所にあり、その内部は、空
気などのガス雰囲気のボイド領域にして使用済燃料集合
体３を収納し測定する。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to FIGS. As shown in FIG. 1, a neutron source intensity measuring device (first device) 1 includes a watertight container on which a neutron absorber 2 is attached and a first predetermined axial position neutron count rate of a spent fuel assembly 3. A neutron detector 5 to be measured and a calculation means 6 for calculating the aggregate neutron source intensity based on a mathematical expression based on the measured value are used. Although not shown, the watertight container is located at the same place as the neutron absorber 2, and the inside thereof is used as a void region of a gas atmosphere such as air, and the spent fuel assembly 3 is housed and measured.
【００１４】また未臨界度測定装置（第２の装置）１１
は、図２に示すように、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、その測定値及び第１の装置で求めた内部中性子源強
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。A subcriticality measuring device (second device) 11
As shown in FIG. 2, is a gray absorber 12 having a weak neutron absorption capacity, and a neutron detector 15 for measuring the second predetermined axial position neutron count rate of the spent fuel assembly 13.
And a calculation means 16 for calculating the collective neutron effective multiplication factor based on a mathematical formula based on the measured value and the internal neutron source intensity axial distribution obtained by the first device.
【００１５】使用済燃料集合体３，１３は、ジルコニウ
ム合金などの被覆管の内部に二酸化ウランなどのペレッ
トを積層した燃料棒が、１本以上、等間隔に配列された
集合体をほぼ等間隔に配置してあり、例えば断面が約２
０cm角で長さ約４ｍの外形形状を有している。燃焼度
は、例えば装荷した核燃料の量に対する燃焼による減損
量の割り合であり、核***によって放出された中性子が
熱中性子に減速され、さらに次の世代に何個の中性子を
生み出すかを表わす量が増倍率Ｋであって、Ｋ＝１が臨
界状態でＫ＜１が未臨界状態を示す。燃焼度の測定は、
原子炉の運転歴データより求めることができるものの、
その解析は極めて繁雑であって、本実施例では以下の手
順により容易に増倍率（未臨界度）を演算することが可
能となる。The spent fuel assemblies 3 and 13 are composed of at least one fuel rod in which pellets such as uranium dioxide are laminated inside a cladding tube such as a zirconium alloy and the fuel rods are arranged at equal intervals. Are located in the
It has an outer shape of 0 cm square and a length of about 4 m. Burnup is, for example, the ratio of the amount of depletion due to combustion to the amount of loaded nuclear fuel, and the amount of neutrons emitted by fission is decelerated into thermal neutrons. In the multiplication factor K, K = 1 indicates a critical state and K <1 indicates a subcritical state. Burnup measurement is
Although it can be obtained from the operational history data of the reactor,
The analysis is extremely complicated, and in this embodiment, the multiplication factor (subcriticality) can be easily calculated by the following procedure.
【００１６】（手順１）図１に示すように、プール水中
に浸した中性子源強度測定装置の第１の装置を用い、中
性子検出器により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を測定する。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。(Procedure 1) As shown in FIG. 1, using a first neutron source intensity measuring apparatus immersed in pool water, a neutron detector is used to measure a first axis at a predetermined cross-sectional position of a spent fuel assembly. The direction predetermined position neutron count rate C 0 is measured. First
The axial predetermined position neutron count C 0 is proportional to the detector position thermal neutron flux calculation value, as shown in Equation 1.
【００１７】[0017]
【数１】 [Equation 1]
【００１８】従来のようにプール水中で測定すると、そ
の測定値は未使用燃料集合体の燃焼度の変動に伴い、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
が変化するため、内部中性子発生率（集合体中性子源強
度Ｓ0＝集合体の所定断面位置における中性子強度）と
第１の軸方向所定位置中性子計数率Ｃ0との相関関係が
複雑なものとなるが、本装置では容器内を空気などのガ
ス雰囲気とすること及び容器の周囲に中性子吸収体を設
けてあるため、燃料部における熱中性子がなくなり、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
がほぼ一定値となり、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とが比例関係となって
容易に集合体中性子源強度Ｓ0を数１により求めること
ができる。When measured in pool water as in the prior art, the measured values are associated with changes in the burnup of the unused fuel assemblies, and the main system assembly neutron effective multiplication factor K 0 and the detector detection efficiency α 0.
Therefore, the correlation between the internal neutron generation rate (aggregate neutron source strength S 0 = neutron intensity at a predetermined cross-sectional position of the aggregate) and the first predetermined axial position neutron count rate C 0 is complicated. However, in this apparatus, since the inside of the container is made to be a gas atmosphere such as air and the neutron absorber is provided around the container, thermal neutrons in the fuel part disappear, and the main system aggregate neutron effective multiplication factor K 0 and Detector detection efficiency α 0
Is a substantially constant value, and the collective neutron source strength S 0 and the first predetermined axial position neutron count rate C 0 have a proportional relationship, and the collective neutron source strength S 0 can be easily obtained by the mathematical expression 1. .
【００１９】（手順２）手順１で求めた集合体中性子源
強度Ｓ0と比例する第１の軸方向所定位置中性子計数率
Ｃ0を基に、図２に示す未臨界度測定装置で計測した第
２の軸方向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。(Procedure 2) Based on the neutron count rate C 0 at the first predetermined axial position, which is proportional to the aggregate neutron source intensity S 0 obtained in procedure 1, the measurement was made by the subcriticality measuring device shown in FIG. The second predetermined axial position neutron count rate C 1 has a relationship with the collective neutron effective multiplication factor K 1 as shown in Formulas 2 and 3.
【００２０】[0020]
【数２】 [Equation 2]
【００２１】[0021]
【数３】 [Equation 3]
【００２２】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（弱い吸収力を有する中
性子吸収体）であり、グレイ吸収体は水平方向の核***
中性子分布を内部中性子源分布と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 (having a weak absorbing power that makes α 1 and A constant values irrespective of the fluctuation of the burnup of the spent fuel assembly). Neutron absorber), the gray absorber is a material that makes the horizontal fission neutron distribution equal to the internal neutron source distribution. A is calculated from the equation 4 by analysis.
【００２３】[0023]
【数４】 [Equation 4]
【００２４】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.
【００２５】[0025]
【発明の効果】本発明によれば、輸送又は貯蔵体系内の
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
度よく測定できるため、適正に燃焼度クレジットを考慮
した臨界安全性を保証することができる。According to the present invention, since the subcriticality of spent fuel assemblies having various burnups in the transportation or storage system can be accurately measured, the criticality safety in which burnup credits are properly considered. Can be guaranteed.
【図１】本発明の一実施例を示す中性子源強度測定装置
の構成図である。FIG. 1 is a configuration diagram of a neutron source intensity measuring apparatus showing an embodiment of the present invention.
【図２】本発明の一実施例を示す未臨界度測定装置の構
成図である。FIG. 2 is a configuration diagram of a subcriticality measuring device showing an embodiment of the present invention.
１ 第１の装置（中性子源強度測定装置） ２ 中性子吸収体 ３ 使用済燃料集合体 ５ 中性子検出器 ６ 演算手段 １１ 第２の装置（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 1 1st apparatus (neutron source intensity measuring apparatus) 2 neutron absorber 3 spent fuel assembly 5 neutron detector 6 calculating means 11 2nd apparatus (subcriticality measuring apparatus) 12 gray absorber 13 spent fuel assembly Body 15 neutron detector 16 computing means
【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１】[Procedure Amendment 1]
【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シス
テム及び未臨界度測定方法Title: Subcriticality measuring system and subcriticality measuring method 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
料集合体の燃焼に伴う中性子実効増倍率の低下を考慮す
る臨界管理（燃焼度クレジット）における未臨界度測定
に係り、特に輸送又は貯蔵体系内の使用済燃料集合体の
未臨界度を測定するのに好適な使用済燃料集合体の未臨
界度測定システム及び未臨界度測定方法に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to subcriticality measurement in criticality control (burnup credit) in consideration of reduction in effective neutron multiplication factor associated with combustion of spent fuel assemblies such as light water reactors, and particularly to transportation or The present invention relates to a subcriticality measuring system and a subcriticality measuring method for a spent fuel assembly suitable for measuring the subcriticality of a spent fuel assembly in a storage system.
【０００２】[0002]
【従来の技術】使用済燃料の取扱い施設及び輸送もしく
は貯蔵用容器（キャスク）の臨界安全設計における未臨
界性の評価では、中性子増倍率が最も高くなるような燃
料を想定する必要がある。従来、この想定において未燃
焼の燃料を前提とした臨界安全設計が行われている。In the evaluation of definitive subcriticality in criticality safety design of the handling facilities and transport or storage container BACKGROUND ART spent fuel (cask), it is necessary to assume the fuel, such as neutron multiplication factor is highest. Conventionally, in this assumption, a criticality safety design based on unburned fuel has been performed.
【０００３】実際には使用済燃料は、燃焼によって実効
的な核***性物質量が減少することにより中性子に対し
て核***する割合が小さくなり、さらに、核***生成物
の蓄積により中性子吸収効果が増大するために、中性子
増倍率が低下している。使用済燃料の取扱い施設の臨界
安全設計及び臨界安全管理において、燃焼に伴う燃料の
中性子増倍率の低下を考慮することを、燃焼度クレジッ
トといい、近年、燃焼度クレジットの考え方を採用する
ことについての技術的検討が世界各国でなされている。
多数の原子力発電所を有する米国では、既に一部の発電
所の燃料貯蔵プールに燃焼度クレジットを取り入れ、貯
蔵容量を増加させている。In practice, spent fuel is burned against neutrons by reducing the effective amount of fissile material.
Ratio of fission Te decreases, further, to the neutron absorption effect increases by the accumulation of fission products, the neutron multiplication factor is decreased. In the criticality safety design and criticality safety management of facilities that handle spent fuel, taking into consideration the decrease in the neutron multiplication factor of fuel due to combustion is called burnup credit, and in recent years, the concept of burnup credit has been adopted. technical studies of One stomach have been made around the world.
In the United States, which has a large number of nuclear power plants, some have already incorporated burnup credits into their fuel storage pools to increase storage capacity.
【０００４】ＡＮＳＩ／ＡＮＳ−５７．７−１９８１
は、使用済燃料貯蔵プールの臨界安全性評価における燃
料の想定を以下の二つの分類のどちらかとしている。ANSI / ANS-57.7-1981
Defines fuel assumptions for criticality safety assessment of spent fuel storage pools as one of the following two categories.
【０００５】ａ）燃焼による核***性物質量の減少は考
慮せず、フール中において中性子増倍率が最も高くなる
ような燃料とすること。A) A fuel that has the highest neutron multiplication factor in the fuel is taken into consideration without considering the decrease in the amount of fissile material due to combustion.
【０００６】ｂ）燃焼による核***性物質量の減少を考
慮する場合には、貯蔵する燃料の最小燃焼度を設定し、
それに対応した中性子増倍率を想定すること、及び実測
により中性子増倍率の低下を確認すること。B) When considering the decrease in the amount of fissile material due to combustion, the minimum burnup of the fuel to be stored is set,
Assuming a neutron multiplication factor corresponding to it, and confirming a decrease in neutron multiplication factor by actual measurement.
【０００７】[0007]
【発明が解決しようとする課題】従来は使用済燃料集合
体の反応度を安全側に高く見積って本体系の臨界解析を
行っているため、臨界安全側へ余裕が大きく、精度に問
題点があった。Conventionally, since the reactivity of the spent fuel assembly is highly estimated on the safe side to perform the criticality analysis of the main body system, there is a large margin on the criticality safe side and there is a problem in accuracy. there were.
【０００８】本発明の目的は、種々の燃焼度を有し、燃
焼度などの運転歴データが不明でも未臨界度の測定を行
うことのできる使用済燃料集合体の未臨界度測定システ
ム及び未臨界度測定方法を提供し、燃焼度クレジットの
導入に寄与ことにある。An object of the present invention is to provide a subcriticality measuring system for a spent fuel assembly which has various burnups and is capable of measuring subcriticality even if operating history data such as burnup is unknown. To provide a method for measuring criticality and contribute to the introduction of burnup credits.
【０００９】[0009]
【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シス
テムは、使用済燃料集合体を収容し該使用済燃料集合体
の燃焼度の変動による検出器検出効率の変化を防止する
工夫を施して第１の軸方向所定位置中性子計数率を測定
し、第１の軸方向所定位置中性子計数率に基づき集合体
中性子源強度を演算する中性子源強度測定装置と、使用
済燃料集合体を収容し使用済燃料集合体の水平方向の核
***中性子分布を内部中性子分布とほぼ同一分布にする
工夫を施して第２の軸方向所定位置中性子計数率を測定
し、第２の軸方向所定位置中性子計数率、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に基づ
き集合体中性子実効増倍率を演算する未臨界度測定装置
とを備えた構成とする。In order to achieve the above object, a subcriticality measuring system for a spent fuel assembly according to the present invention stores a spent fuel assembly and burns the spent fuel assembly. Measure the 1st axial predetermined position neutron count rate by devising to prevent the change of detector detection efficiency due to the fluctuation of the degree, and calculate the aggregate neutron source intensity based on the 1st axial predetermined position neutron count rate Neutron source intensity measuring device for storing the spent fuel assemblies and devising a device to make the horizontal fission neutron distribution of the spent fuel assemblies almost the same as the internal neutron distribution. Counting rate is measured, second predetermined axial position neutron counting rate, aggregate neutrons
A structure in which a subcriticality measuring device for calculating a first axial position neutron count rate based Dzu <br/> can aggregate effective neutron multiplication factor of the source intensity.
【００１０】そして使用済燃料集合体の未臨界度測定方
法においては、使用済燃料集合体の燃焼度の変動による
検出器検出効率の変化を防止する工夫を施して第１の軸
方向所定位置中性子計数率を測定し、第１の軸方向所定
位置中性子計数率に基づき集合体中性子源強度を演算す
る第１の手順と、使用済燃料集合体の水平方向の核***
中性子分布を内部中性子分布とほぼ同一分布にする工夫
を施して第２の軸方向所定位置中性子計数率を測定し、
第２の軸方向所定位置中性子計数率、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に基づき集
合体中性子実効増倍率を演算する第２の手順とよりなる
構成とする。In the method for measuring the subcriticality of the spent fuel assembly, the burnup of the spent fuel assembly is changed.
The first axial direction predetermined position neutron counting rate is measured by devising a device for preventing a change in detector detection efficiency , and the aggregate neutron source intensity is calculated based on the first predetermined axial position neutron counting rate. Measure the second axial predetermined position neutron count rate by applying a procedure and a device to make the horizontal fission neutron distribution of the spent fuel assembly almost the same as the internal neutron distribution,
Second predetermined axial position neutron count rate, aggregate neutron source strength
The second procedure for calculating the effective neutron multiplication factor of the aggregate neutrons based on the neutron count rate of the first predetermined position in the axial direction related to the degree .
【００１１】[0011]
【作用】本発明によれば、中性子源強度測定装置では、
使用済燃料集合体が空気等のガス雰囲気中で水密内容器
に設置されるため体系内で中性子の熱化が生じないこ
と、かつ容器の外周に中性子吸収体が設けられ、装置が
水中に浸されているため、中性子が使用済燃料集合体の
外周の水部により反射されて容器に入る熱中性子が吸収
されることにより、したがって燃料部の熱中性子がなく
なり、本体系集合体中性子実効増倍率及び検出器検出効
率がほぼ一定値となる。このため各軸方向の集合体内部
中性子発生率（集合体内部中性子源強度）と第１の軸方
向所定位置中性子計数率とが比例関係を有するようにな
り、容易に第１の軸方向所定位置中性子計数率が測定さ
れ、測定値より集合体中性子源強度及び軸方向分布が演
算される。According to the present invention, in the neutron source intensity measuring device,
Since the spent fuel assembly is installed in a watertight inner container in a gas atmosphere such as air, neutron thermalization does not occur in the system, and a neutron absorber is installed on the outer periphery of the container so that the device can be immersed in water. because they are, by the neutron thermal neutrons entering the vessel is reflected by the water portion of the outer periphery of the spent fuel assemblies is absorbed, thus eliminating the thermal neutrons in the fuel unit, main system assembly neutron effective multiplication factor And the detection efficiency of the detector becomes almost constant. Therefore, the aggregate internal neutron generation rate (internal aggregate neutron source intensity) in each axial direction and the neutron count rate at the first axial predetermined position have a proportional relationship, and the first axial predetermined position can be easily obtained. The neutron count rate is measured, and the aggregate neutron source intensity and axial distribution are calculated from the measured values.
【００１２】つぎに未臨界度測定装置では、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***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。Next, in the subcriticality measuring device, the gray absorber makes the internal neutron source distribution equal to the horizontal fission neutron distribution induced by the neutrons.
The detector detection efficiency due to the change in burnup becomes almost constant, and the detector detection efficiency is calculated in advance for unirradiated fuel, etc., and based on the neutron count rate of the first and second axial predetermined positions. The effective neutron multiplication factor (subcriticality) is calculated accurately.
【００１３】[0013]
【実施例】本発明の一実施例を図１及び図２を参照しな
がら説明する。図１に示すように、中性子源強度測定装
置（第１の装置）１は、中性子吸収体２を張付けた水密
容器と、使用済燃料集合体３の第１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５と、その測定値を
基に数式に基づいて集合体中性子源強度を演算する演算
手段６とよりなる構成とする。水密容器は、図示しない
が中性子吸収体２と同一の場所にあり、その内部は、空
気などのガス雰囲気のボイド領域にして使用済燃料集合
体３を収納し測定する。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to FIGS. As shown in FIG. 1, a neutron source intensity measuring device (first device) 1 includes a watertight container on which a neutron absorber 2 is attached and a first predetermined axial position neutron count rate of a spent fuel assembly 3. A neutron detector 5 to be measured and a calculation means 6 for calculating the aggregate neutron source intensity based on a mathematical expression based on the measured value are used. Although not shown, the watertight container is located at the same place as the neutron absorber 2, and the inside thereof is used as a void region of a gas atmosphere such as air, and the spent fuel assembly 3 is housed and measured.
【００１４】また未臨界度測定装置（第２の装置）１１
は、図２に示すように、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、その測定値及び第１の装置で求めた軸方向所定位置
中性子計数率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。A subcriticality measuring device (second device) 11
As shown in FIG. 2, is a gray absorber 12 having a weak neutron absorption capacity, and a neutron detector 15 for measuring the second predetermined axial position neutron count rate of the spent fuel assembly 13.
And its measured value and the predetermined axial position determined by the first device
The calculation means 16 calculates the effective neutron multiplication factor of the aggregate based on a mathematical expression based on the neutron count rate .
【００１５】使用済燃料集合体３，１３は、ジルコニウ
ム合金などの被覆管の内部に二酸化ウランなどのペレッ
トを積層した燃料棒が、ほぼ等間隔に配置してあり、例
えば断面が約２０cm角で長さ約４ｍの外形形状を有して
いる。燃焼度は、例えば装荷した核燃料の量に対する燃
焼による減損量の割合であり、核***によって放出され
た１個の中性子が熱中性子に減速され、さらに次の世代
に何個の中性子を生み出すかを表わす量が増倍率Ｋであ
って、Ｋ＝１が臨界状態でＫ＜１が未臨界状態を示す。
燃焼度の測定は、原子炉の運転歴データより求めること
ができるものの、その解析は極めて繁雑であって、本実
施例では以下の手順により容易に増倍率（未臨界度）を
演算することが可能となる。The spent fuel assemblies 3, 13, inside the fuel rods by laminating a pellet, such as uranium dioxide of the cladding such as zirconium alloy, Yes disposed in almost equal intervals, for example, cross section of about 20cm square And has an outer shape of about 4 m in length. Burnup is, for example, a percentage of the impairment amount by combustion to the amount of loading the nuclear fuel, or one neutron emitted by fission is decelerated to thermal neutrons, further produce many neutrons to the next generation The indicated amount is a multiplication factor K, K = 1 indicates a critical state and K <1 indicates a subcritical state.
Although the burnup can be measured from the operation history data of the reactor, the analysis is extremely complicated, and in this embodiment, the multiplication factor (subcriticality) can be easily calculated by the following procedure. It will be possible.
【００１６】（手順１）図１に示すように、プール水中
に浸した中性子源強度測定装置の第１の装置を用い、中
性子検出器により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を測定する。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。(Procedure 1) As shown in FIG. 1, using a first neutron source intensity measuring apparatus immersed in pool water, a neutron detector is used to measure a first axis at a predetermined cross-sectional position of a spent fuel assembly. The direction predetermined position neutron count rate C 0 is measured. First
The axial predetermined position neutron count C 0 is proportional to the detector position thermal neutron flux calculation value, as shown in Equation 1.
【００１７】[0017]
【数１】 [Equation 1]
【００１８】従来のようにプール水中で測定すると、そ
の測定値は未使用燃料集合体の燃焼度の変動に伴い、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
が変化するため、内部中性子発生率（集合体中性子源強
度Ｓ0＝集合体の所定断面位置における中性子源強度）
と第１の軸方向所定位置中性子計数率Ｃ0との相関関係
が複雑なものとなるが、本装置では容器内を空気などの
ガス雰囲気とすること及び容器の周囲に中性子吸収体を
設けてあるため、燃料部における熱中性子がなくなり、
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がほぼ一定値となり、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とが比例関係となっ
て容易に集合体中性子源強度Ｓ0を数１により求めるこ
とができる。When measured in pool water as in the prior art, the measured values are associated with changes in the burnup of the unused fuel assemblies, and the main system assembly neutron effective multiplication factor K 0 and the detector detection efficiency α 0.
Order to make the transition, internal neutron generation rate (neutron source intensity at a given cross-sectional position of the aggregate neutron source strength S 0 = aggregate)
And the first predetermined axial position neutron count rate C 0 have a complicated correlation. However, in this device, a gas atmosphere such as air is provided in the container and a neutron absorber is provided around the container. Because there is no thermal neutron in the fuel part,
This system aggregate effective neutron multiplication factor K 0 and detector detection efficiency α 0 become almost constant values, and the aggregate neutron source intensity S 0 and the first
The neutron count rate C 0 of the axial predetermined position has a proportional relationship, and the aggregate neutron source intensity S 0 can be easily obtained by the formula 1.
【００１９】（手順２）手順１で求めた集合体中性子源
強度Ｓ0と比例する第１の軸方向所定位置中性子計数率
Ｃ0を基に、図２に示す未臨界度測定装置で計測した第
２の軸方向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。(Procedure 2) Based on the neutron count rate C 0 at the first predetermined axial position, which is proportional to the aggregate neutron source intensity S 0 obtained in procedure 1, the measurement was made by the subcriticality measuring device shown in FIG. The second predetermined axial position neutron count rate C 1 has a relationship with the collective neutron effective multiplication factor K 1 as shown in Formulas 2 and 3.
【００２０】[0020]
【数２】 [Equation 2]
【００２１】[0021]
【数３】 [Equation 3]
【００２２】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（弱い吸収力を有する中
性子吸収体）であり、グレイ吸収体は水平方向の核***
中性子分布を内部中性子源分布と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 (having a weak absorbing power that makes α 1 and A constant values irrespective of the fluctuation of the burnup of the spent fuel assembly). Neutron absorber), the gray absorber is a material that makes the horizontal fission neutron distribution equal to the internal neutron source distribution. A is calculated from the equation 4 by analysis.
【００２３】[0023]
【数４】 [Equation 4]
【００２４】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.
【００２５】[0025]
【発明の効果】本発明によれば、輸送又は貯蔵体系内の
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
度よく測定できるため、適正に燃焼度クレジットを考慮
した臨界安全管理手法の確立に寄与することができる。According to the present invention, since the subcriticality of spent fuel assemblies having various burnups in the transportation or storage system can be accurately measured, the criticality safety control in which burnup credits are properly taken into consideration. It can contribute to the establishment of the method .
【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中性子源強度測定装置
の構成図である。FIG. 1 is a configuration diagram of a neutron source intensity measuring apparatus showing an embodiment of the present invention.
【図２】本発明の一実施例を示す未臨界度測定装置の構
成図である。FIG. 2 is a configuration diagram of a subcriticality measuring device showing an embodiment of the present invention.
【符号の説明】 １ 第１の装置（中性子源強度測定装置） ２ 中性子吸収体 ３ 使用済燃料集合体 ５ 中性子検出器 ６ 演算手段 １１ 第２の装置（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[Explanation of Codes] 1 first device (neutron source intensity measuring device) 2 neutron absorber 3 spent fuel assembly 5 neutron detector 6 computing means 11 second device (subcriticality measuring device) 12 gray absorber 13 spent fuel assembly 15 neutron detector 16 computing means
Claims (2)
集合体の燃焼度の変動による係数率の変化を防止する工
夫を施して第１の軸方向所定位置中性子計数率を測定
し、該第１の軸方向所定位置中性子計数率に基づき集合
体中性子源強度を演算する中性子源強度測定装置と、前
記使用済燃料集合体を収容し該使用済燃料集合体の水平
方向の核***中性子分布を内部中性子分布とほぼ同一分
布にする工夫を施して第２の軸方向所定位置中性子計数
率を測定し、該第２の軸方向所定位置中性子計数率、前
記第１の軸方向所定位置中性子計数率及び前記集合体中
性子源強度に基づき集合体中性子実効増倍率を演算する
未臨界度測定装置とを備えたことを特徴とする使用済燃
料集合体の未臨界度測定システム。1. A first axial predetermined position neutron count rate is measured by accommodating a spent fuel assembly to prevent a change in a coefficient rate due to a change in burnup of the spent fuel assembly, A neutron source intensity measuring device for calculating an aggregate neutron source intensity based on the first predetermined axial position neutron count rate, and a horizontal fission neutron distribution of the spent fuel assembly containing the spent fuel assembly. To measure the second predetermined axial position neutron count rate, and the second predetermined axial position neutron count rate, the first predetermined axial position neutron count A subcriticality measuring system for a spent fuel assembly, comprising: a subcriticality measuring device for calculating an effective neutron multiplication factor of the aggregate based on the rate and the aggregate neutron source intensity.
係数率の変化を防止する工夫を施して第１の軸方向所定
位置中性子計数率を測定し、該第１の軸方向所定位置中
性子計数率に基づき集合体中性子源強度を演算する第１
の手順と、前記使用済燃料集合体の水平方向の核***中
性子分布を内部中性子分布とほぼ同一分布にする工夫を
施して第２の軸方向所定位置中性子計数率を測定し、該
第２の軸方向所定位置中性子計数率、前記第１の軸方向
所定位置中性子計数率及び前記集合体中性子源強度に基
づき集合体中性子実効増倍率を演算する第２の手順とよ
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
定方法。2. A first axial predetermined position neutron count rate is measured by devising a device for preventing a change in coefficient rate due to a change in burnup of a spent fuel assembly, and the first axial predetermined position neutron count rate is measured. First to calculate aggregate neutron source intensity based on count rate
And the horizontal axis fission neutron distribution of the spent fuel assembly is made to be almost the same as the internal neutron distribution, and the second predetermined axial position neutron count rate is measured, and the second axis Direction predetermined position neutron count rate, the first axial direction predetermined position neutron count rate, and a second procedure for calculating an aggregate neutron effective multiplication factor based on the aggregate neutron source intensity Fuel assembly subcriticality measurement method.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4215039A JPH0659085A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring system for spent fuel assembly and non-criticality measurement |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4215039A JPH0659085A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring system for spent fuel assembly and non-criticality measurement |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH0659085A true JPH0659085A (en) | 1994-03-04 |
Family
ID=16665749
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP4215039A Withdrawn JPH0659085A (en) | 1992-08-12 | 1992-08-12 | Non-criticality measuring system for spent fuel assembly and non-criticality measurement |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0659085A (en) |
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2005338042A (en) * | 2004-05-31 | 2005-12-08 | Toshiba Corp | Critical safety design program for equipments for transporting and storing spent fuel |
| JP2007285987A (en) * | 2006-04-20 | 2007-11-01 | Toshiba Corp | Device for accommodating fuel assembly and method for positioning system for measuring subcriticality |
| JP2008039509A (en) * | 2006-08-03 | 2008-02-21 | Toshiba Corp | Method for loading irradiated fuel into subcritical neutron multiplication system and method for calculating effective multiplication factor of irradiated fuel |
| JP2009168801A (en) * | 2007-12-20 | 2009-07-30 | Westinghouse Electric Co Llc | Method for improving combustion deduction of spent nuclear fuel |
| CN104167232A (en) * | 2014-08-19 | 2014-11-26 | 中兴能源装备有限公司 | Dry-type spent fuel storage device |
| CN105261402A (en) * | 2015-09-07 | 2016-01-20 | 中广核工程有限公司 | Fuel storage tank for nuclear power plant spent fuel dry-type storage |
| JP2016024154A (en) * | 2014-07-24 | 2016-02-08 | 日立Ｇｅニュークリア・エナジー株式会社 | Estimation method of subcritical state and subcritical state estimation system |
| CN106448783A (en) * | 2016-12-12 | 2017-02-22 | 深圳中广核工程设计有限公司 | Dual-functional spent-fuel storage and transportation metal tank of nuclear power plant |
-
1992
- 1992-08-12 JP JP4215039A patent/JPH0659085A/en not_active Withdrawn
Cited By (9)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2005338042A (en) * | 2004-05-31 | 2005-12-08 | Toshiba Corp | Critical safety design program for equipments for transporting and storing spent fuel |
| JP2007285987A (en) * | 2006-04-20 | 2007-11-01 | Toshiba Corp | Device for accommodating fuel assembly and method for positioning system for measuring subcriticality |
| JP2008039509A (en) * | 2006-08-03 | 2008-02-21 | Toshiba Corp | Method for loading irradiated fuel into subcritical neutron multiplication system and method for calculating effective multiplication factor of irradiated fuel |
| JP2009168801A (en) * | 2007-12-20 | 2009-07-30 | Westinghouse Electric Co Llc | Method for improving combustion deduction of spent nuclear fuel |
| JP2016024154A (en) * | 2014-07-24 | 2016-02-08 | 日立Ｇｅニュークリア・エナジー株式会社 | Estimation method of subcritical state and subcritical state estimation system |
| CN104167232A (en) * | 2014-08-19 | 2014-11-26 | 中兴能源装备有限公司 | Dry-type spent fuel storage device |
| CN105261402A (en) * | 2015-09-07 | 2016-01-20 | 中广核工程有限公司 | Fuel storage tank for nuclear power plant spent fuel dry-type storage |
| CN106448783A (en) * | 2016-12-12 | 2017-02-22 | 深圳中广核工程设计有限公司 | Dual-functional spent-fuel storage and transportation metal tank of nuclear power plant |
| WO2018108073A1 (en) * | 2016-12-12 | 2018-06-21 | 深圳中广核工程设计有限公司 | Nuclear power plant spent fuel storage and transportation metal tank |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US4325785A (en) | Method and apparatus for measuring the reactivity of a spent fuel assembly | |
| JPH0659085A (en) | Non-criticality measuring system for spent fuel assembly and non-criticality measurement | |
| JP5752467B2 (en) | Reactor fuel non-destructive burnup evaluation method and apparatus | |
| KR101515638B1 (en) | Method of improving the spent nuclear fuel burnup credit | |
| JP3708599B2 (en) | Subcriticality evaluation method when storing spent fuel assemblies | |
| JP2542883B2 (en) | Effective multiplication factor measurement method for subcritical systems loaded with irradiation fuel | |
| Mihalczo et al. | 252 Cf-source-driven neutron noise analysis method | |
| JP2012122929A (en) | Method and apparatus for evaluating nondestructive burnup of reactor fuel | |
| US4881247A (en) | Measuring nuclear fuel burnup | |
| You et al. | Introduction of burn-up credit in nuclear criticality safety analysis | |
| JPH01199195A (en) | Effective multiplication factor measuring method of irradiation fuel charged subcritical system | |
| JP4664645B2 (en) | Method for measuring neutron emission rate of irradiated fuel assemblies | |
| JP4723966B2 (en) | Np content inspection method and Np content inspection device | |
| JPH11118981A (en) | Non-critical measurement system of spent nuclear fuel system | |
| Usami et al. | Measurement of shielding characteristics in the prototype FBR Monju | |
| JPS6155075B2 (en) | ||
| JPH05288888A (en) | Measuring method of effective multiplication factor and disposing method of neutron detector | |
| JP2004077291A (en) | Burnup measuring device for spent fuel | |
| JP2015010893A (en) | Criticality safety management method | |
| JPH0659089A (en) | Transport container for spent fuel assembly | |
| Radulescu et al. | Review of Experimental Data for Validating Computer Codes Used in Shielding Calculations for Spent Fuel Storage and Transportation Systems | |
| JPH0226754B2 (en) | ||
| Mihalczo et al. | Feasibility of subcriticality and NDA measurements for spent fuel by frequency analysis techniques with 252 Cf | |
| JP2023031564A (en) | Criticality management method for used mixed oxide fuel and used mixed oxide fuel criticality safety evaluation device | |
| JPH0659082A (en) | Burnup degree credit introduction for spent fuel assembly |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| A300 | Withdrawal of application because of no request for examination |
Free format text: JAPANESE INTERMEDIATE CODE: A300 Effective date: 19991102 |