Reactor Neutron Capture Cross Section of 60.3-day 124Sb

The radioactive nuclides ¹²⁴Sb (T _1/2=60.3d) and ¹²⁵Sb (T_1/2=2.77yr) were produced from natural antimony by JRR-3 reactor irradiation of 283.5 h through the single and double capture processes. After cooling of 3.50 yr, the γ-ray spectrum of the antimony sample irradiated was measured by a 50 cc coaxial type Ge(Li) detector, and the photo-peak yield ratio of ¹²⁵Sb (E _r=428keV) to ¹²⁴Sb (E _r=1.691 MeV) was obtained. By using a relation between this photo-peak yield ratio and the ¹²⁴Sb (n, γ) ¹²⁶Sb cross section, the reactor neutron capture cross section of 60.3-day ¹²⁴Sb was obtained as 17.4:5:^+2.8 _?2.5b. The thermal neutron flux at the position of antimony sample irradiated was estimated as (4.92±0.38) ×10¹²n/cm²·s by measuring the 1.333-MeV photo-peak yield of ⁶⁰Co, which was activated by reactor irradiation of cobalt impurity contained in the antimony sample.     

Vaiidition of ORIGEN Computer Code by Measurements on Nuclear Fuels of JPDR-1

The depletion and production amounts of U, Pu, transplutonium nuclides and fission products (FPs) measured on the fuel of JPDR-1 were corrected to take account of the performance history of irradiation and cooling using the results of three-dimensional nuclear-thermo-hydrodynamic and nuclide depletion and production calculations. Except a few nuclides, the corrected values proved to agree well with the values calculated by the ORIGEN computer code. Further enhancement of calculational accuracy calls for systematic re-evaluation of neutron cross sections on the basis of neutron spectrum in nuclear fuel.     

Existence of the fifth unstable nuclided series

ＥｘｉｓｔｅｎｃｅｏｆｔｈｅｆｉｆｔｈｕｎｓｔａｂｌｅｎｕｃｌｉｄｅｄｓｅｒｉｅｓＺｈａｎｇＪｉａ－Ｈｕａ（张家骅）（ＳｈａｎｇｈａｉＩｎｓｔｉｔｕｔｅｏｆＮｕｃｌｅａｒＲｅｓｅａｒｃｈ，ｔｈｅＣｈｉｎｅｓｅＡｃａｄｅｍｙｏｆＳｃｉｅｎｃｅｓ，Ｓｈａ...     

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of ²³⁷Np, ²⁴¹Am and ²⁴³Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 10¹⁴ to 10¹⁷ n/cm².s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n _net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n _net value in the fast neutron field was positive (0.0–1.0) for ²³⁷Np, ²⁴¹Am, ²⁴³Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n _net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 10¹⁴ n/cm².s and 0.0 at 10¹⁶ n/cm².s. In the high flux region of 10¹⁶ n/cm².s, TRU in the high-level waste can be transmuted by fission without consuming additional neutrons. In the flux region about 10¹⁴ n/cm².s, the transmutation of TRU in the high-level waste by fission requires about one neutron.     

Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10¹⁷, 6.1 × 10¹⁶ (E < 0.55 eV), 7.6 × 10¹⁶ (E > 0.1 MeV) and 4.0 × 10¹⁶ (E > 1 MeV) m⁻² s⁻¹. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.     

第五不稳定核素系及其特性

将４个分别以２３８Ｕ、２３２Ｔｈ、２３５Ｕ和２３７Ｎｐ为母体的，由于内因的放射性衰变作用而形成的不稳定核素系加以剖析，并分析其成系所遵循的规律，发育阶段和进入平衡阶段后所显出的特性等。以此作为参照样本，分析了２３８Ｕ（或２３２Ｔｈ）在核反应堆的中子通量连续作用下所产生的各代衍生核素的演化过程，并得出完全符合形成不稳定核素系的条件。为此特称之为第五不稳定核素系。它在达到平衡阶段时所显现的特征为：全系核素在外因作用下（即堆中子的作用下）所导致的原子损耗几乎全部转化为裂变产物。     

针对裂变产额和半衰期的燃耗计算灵敏度和不确定度分析方法

基于广义微扰理论推导了裂变产额和半衰期的燃耗灵敏度系数理论模型,该模型考虑了原子核密度和中子通量的相互影响,并开发了燃耗计算中有效增殖因数和原子核密度等响应参数对核数据的灵敏度和不确定度分析程序。基于评价核数据中裂变产物独立产额的标准差数据,产生了针对压缩燃耗数据库的裂变产额协方差矩阵,以提高不确定度的计算精度。基于ENDF/B-Ⅶ.1数据库量化了UAM基准题TMI-1栅元无限增殖因数及重要裂变产物和重核的原子核密度由裂变产额和半衰期引入的不确定度。数值结果表明,对于栅元无限增殖因数,裂变产额和半衰期引入的不确定度很小;对于部分裂变产物的原子核密度,裂变产额和半衰期会引入较大的不确定度。     

Burnup Calculation Sensitivity and Uncertainty Analysis Method of Fission Yield and Decay Half-life

The calculation model of sensitivity coefficient for decay half-life and fission product yield in burnup calculation was derived based on generalized perturbation theory, which considered the interaction between nuclear concentration and neutron flux. A code was developed to calculate sensitivity and uncertainty of effective neutron multiplication factors and nuclide concentration caused by nuclear data. Covariance matrix of fission yield for a simplified burnup library was generated based on standard deviation data of independent fission yield in evaluated nuclear data library to improve the accuracy of uncertainty quantification. Uncertainties induced by decay half-life and fission yield on infinite neutron multiplication factors and nuclide concentration for TMI-1 pin-cell in the UAM burnup benchmark were quantified based on ENDF/B-Ⅶ.1. The numerical results show that the uncertainty of infinite neutron multiplication factors induced by decay half-lives and fission yields is low, while the uncertainty of concentration of some fission product nuclide is high.     

Characteristics of Neutron Guide Tubes in Upgraded JRR-3

Abstract

The JRR-3 has been upgraded to be a new high performance research reactor JRR-3M with neutron guide tubes on a large scale and a cold neutron source. The neutron fluxes and spectra were measured at the end of the two thermal and three cold neutron guide tubes. The gain of the cold neutron source is also found from these spectra. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength 2 Å are 1.2x10⁸ n/cm².s at a reactor power of 20 MW. The neutron fluxes of cold guide tubes are 2.0x 10⁸ n/cm².s with characteristic wavelength 4 Å and 1.4x10⁸ n/cm².s with 6 A when the cold neutron source is operated. The neutron spectra measured by the time-of-flight method agree well with their designed ones. The gains of the cold neutron source are 8 for 4 Å and 20 for 6 Å at a reactor power of 20 MW.     

Calculation of Beta-Ray Spectra from Individual and Aggregate Fission Products

The β-ray spectra of individual fission products were calculated by using the β-decay data assuming every β-decay to be allowed transition. For the nuclides without measured decay data the β-feeding function was evaluated with the gross theory of β-decay and the β-ray spectrum was calculated from the function. The measured decay data were also supplemented with the data calculated by the gross theory for the excitation energy range above the highest measured excitation energy level. The β-ray spectra from aggregate fission products after a burst fission were calculated by using the β-ray spectrum and the atom number of each fission product nuclide and they were compared with the ones measured for thermal neutron induced fission of ²³⁵U, ²³⁹Pu and ²⁴¹Pu at Oak Ridge National Laboratory. The spectrum calculations showed excellent agreement with the measured data at shorter cooling times than 10s when many short-lived nuclides without measured decay data contributed considerably to the spectrum.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc…). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called “BUCAL1”. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k_∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.     

用于硼中子俘获治疗的医院中子照射器的反应堆光致缓发中子参数

在介绍单群扩散方程基础上,引入堆芯和反射层的中子价值,根据考虑了光致缓发中子及其价值因素的点堆动态方程,建立了利用现有计算程序进行计算和分析的方法,分析了医院中子照射器光致缓发中子的特性参数,在原有6组缓发中子基础上增加了9组光致缓发中子,为进一步进行用于硼中子俘获治疗的医院中子照射器反应堆的点堆动力学研究提供了重要参数。     

Emission Probabilities of Gamma Rays from the Decay of 233Pa and 238Np,and the Thermal Neutron Capture Cross Section of 237Np

In order to determine the thermal neutron capture cross section of ²³⁷Np, the relevant γ emission probabilities of the 312-keV γ-ray from the decay of ²³³Pa and the 984-keV γ-ray from the decay of ²³⁸Np are deduced from the ratio of the emission rate to the activity. The emission rate and activity are measured with a Ge detector and a Si detector, respectively. The measured emission probability for 312-keV γ-ray is 41.6±0.9% and that for 984-keV γ-ray is 25.2±0.5%. The emission probabilities are used to correct the thermal neutron capture cross section of ²³⁷Np reported previously, and gives 168±6b. The neutron capture cross section is also determined as 169±6b by α-ray spectroscopic method. The measured emission probabilities and capture cross section are compared with others from references. By averaging these values deduced by different methods, the value of 169±4b is recommended as the thermal neutron capture cross section of ²³⁷Np for 2,200 m/s neutrons.     

CRAM在放射性核素存量计算中的应用

在反应堆中,组成材料的稳定核素经受强中子辐照后,会被活化成放射性核素。这些核素及其衰变产物对工作人员的职业辐照剂量具有重要贡献。为了更好地进行人员的辐射防护工作,需要对放射性核素的存量进行精确计算。相对于核素平衡方程的其它求解方法,切比雪夫有理逼近方法(Chebyshev Rational Approximation Method,CRAM)在计算精度和效率方面具有综合性优势。首先介绍了CRAM的基本理论,随后选取典型的例题进行了测试验证。与解析解对比的结果表明,采用CRAM进行中子辐照下的核素活化衰变计算能够取得不错的效果,但是用于核素长期衰变计算可能导致计算错误。针对此问题,将收缩乘方技术与CRAM相结合,取得了正确的计算结果,拓展了CRAM的适用范围。     

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Comparing with the fission product nuclide (FP) decay heat summation calculation result in MeV/sec/fission based on the JENDL FP decay and yield data files 2011 for the burst fission, FP decay heat calculated by ORIGEN2.2 coupled with JENDL-4.0 base library ORLIBJ40 was verified at the cooling time from 1 sec to 10⁸ sec for ²³⁵U (thermal), ²³⁸U (fast), ²³⁹Pu (thermal) and ²⁴¹Pu (thermal). For these fission nuclides, FP decay heat calculated by CASMO5 at the same cooling time after a short irradiation (10⁴ sec) was also compared with that of ORIGEN2.2. In the analysis of decay heat measurements at the cooling time from 2.3 years to 27 years consisting of four data sets on the fuel assemblies discharged from the US PWRs and BWRs, and the Swedish PWRs and BWRs, the average values of the ratios of the calculated to measured results (C/E's) were from 0.972 to 1.031 for ORIGEN2.2, and from 0.977 to 1.016 for CASMO5. The standard deviations of C/E's for the four data sets were from 0.02 to 0.03 for the both codes except for those of the US BWR fuel assemblies which were from 0.11 to 0.12. The obtained C/E's were similar to those in the precedent study.     

Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (I)

Fission gas release from a UO²-graphite mixture was studied during irradiation with the use of the Fission Gas Release Loop in the JRR-3 reactor. The release rates of fission krypton and xenon increased proportionally with neutron flux (6×10¹⁰–6×10¹² n/cm²·sec) and exponentially with temperature (400°–1,000°C). A burst of fission gas was observed when the specimen was abruptly heated to a higher temperature. These results can be explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission fragments and released through annealing of the defects.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

辐照对银铟镉控制棒价值的影响分析

为分析银铟镉（Ag-In-Cd）控制棒内各核素经反应堆中子辐照后的消耗情况以及核素消耗对控制棒价值的影响，本研究采用蒙特卡罗程序模拟了Ag-In-Cd控制棒内主要核素在反应堆运行期间的燃耗，并结合控制棒宏观中子吸收截面和控制棒内的中子注量率水平变化，分析了辐照前后控制棒价值的变化。研究结果表明，控制棒中113Cd随着辐照时间增加而加速消耗，107Ag、109Ag和115In消耗速率相对较慢；控制棒总的宏观中子吸收截面在辐照后降低，但是107Ag、109Ag和115In的中子吸收截面明显地增加；辐照后控制棒内的中子注量率增大，控制棒总中子吸收率无明显变化，即控制棒价值无明显变化。