Neutron economy and nuclear data for transmutation of long-lived fission products

In the study of Self-Consistent Nuclear Energy System, the following 29 long-lived fission products (LLFPs) have been selected to be transmuted into stable or short-lived nuclides: ¹⁰⁶Ru, ¹⁰²Rh, ¹⁰⁹Cd, ¹²⁵Sb, ¹³⁴Cs, ^146,147Pm, ^154,155Eu, ¹⁷¹Tm, ⁸⁵Kr, ⁹⁰Sr, ^93mNb, ^113mCd, ^121mSn, ¹³⁷Cs, ¹⁵¹Sm, ¹⁵²Eu, ^108mAg, ¹⁵⁸Tb, ^166mHo, ⁷⁹Se, ⁹³Zr, ⁹⁴Nb, ⁹⁹Tc, ¹⁰⁷Pd, ¹²⁶Sn, ¹²⁹I, ¹³⁵Cs. In the present study, the number of neutrons necessary for the transmutation of the 29 LLFPs with an FBR was evaluated, and the present status of the (n, γ) and (n,2n) cross section data of the 29 LLFPs in JENDL-3.2 and ENDF/B-VI was investigated. The main results of the present study are as follows: (1)only 0.25 neutron per fission is necessary for the transmutation of the 29 LLFPs with isotopic separation, whereas 6.8 neutrons are necessary with chemical separation, (2)the accuracy of the cross sections is 30 to 100% except for the (n, γ) cross sections of limited nuclides in limited incident neutron energy regions.     

The determination of the Nb(n,n′) Nb cross-section in the neutron energy range 1–6 MeV

Measurements of the 30 keV isomer production cross-section in ⁹³Nb, excited by neutron inelastic scattering have been made at 10 neutron energies in the range 1–6 MeV. Small foils of niobium and suitable monitor elements were exposed to high fluences of monoenergetic neutrons and the cross-section was determined from the K X-ray activities so induced. The high purity Ge detector used in these X-ray measurements was calibrated against a standard solution of ^93mNb which is used as an international reference material for reactor neutron dosimetry. The neutron fluence of the irradiation was measured with a low-efficiency ²³⁵U fission chamber in which the thin fissile deposit was located immediately behind the niobium and monitor foils. The latter acted as secondary measures of the neutron fluence but could be used as the primary standard in the event of failure of the fission chamber. Thus, the production cross-section was measured relative to the ²³⁵U fission cross-section which is a standard reference cross-section. Comparisons are made with other experimental data and with nuclear model calculations of the cross-section and recommended values based on these and our experimental data are presented.     

The cross section for the K(n,p)Ar reaction between 14.2 and 17.2 MeV

Activation techniques have been used to measure the cross section for the ⁴¹K(n,p)⁴¹Ar reaction between 14.2 and 17.2 MeV. Neutrons were produced by the ³H(d,n)⁴He reaction, and the mixed-power method was used to measure the neutron flux through the ²⁷Al(n,)²⁴Na reaction. The activated samples were counted for the 1294 keV, 1.827 h γ-activity of ⁴¹Ar and the 1369 keV, 15.03 h γ-activity of ²⁴Na using a 16% Ge(Li) detector and a 4096-channel analyzer. The cross sections for the ⁴¹K(n,p)⁴¹Ar reaction using the mixed-power method were found to be 53 ± 3 mbarn at 14.2 ± 0.2 MeV, 47 ± 3 mbarn at 15.2 ± 0.2 MeV, 41 ± 3 mbarn at 16.2 ± 0.2 MeV and 36 ± 4 mbarn at 17.2 ± 0.2 MeV. The associated-particle method was also used for measuring the neutron flux in order to check the mixed-powder result at 14.2 MeV. The average cross section for three associated-particle runs at 14.2 MeV was found to be 50 ± 3 mbarn which, within experimental error, agrees with the mixed-powder value.     

中子俘获反应截面在线测量技术研究

利用γ全吸收型4π BaF₂探测装置,对中子俘获反应截面进行了在线测量。基于HI-13串列加速器提供的脉冲化质子束,通过⁷Li(p,n) ⁷Be反应产生中子,构建了keV能区中子源实验条件,经屏蔽准直后的中子轰击样品,应用4π BaF₂装置在线测量(n,γ)反应复合核退激时释放的瞬发γ射线级联,测量了Au、C、Nb、空白等样品。通过计算⁹³Nb(n,γ)⁹⁴Nb和¹⁹⁷Au(n,γ)¹⁹⁸Au两个反应的截面数据比值并与文献数据比对,检验了4π BaF₂探测装置和(n,γ)反应截面在线测量技术,为在中国散裂中子源（CSNS）上顺利开展(n,γ)反应截面数据测量工作提供了技术支持。     

贫化铀球装置内的~(238)U(n,2n)反应率实验研究

采用两套不同尺寸的贫化铀球装置开展了装置内部的238 U(n,2n)反应率实验研究,利用PD-300加速器D-T中子源辐照实验装置,源强变化采用伴随粒子法监测,238 U圆片放置在实验装置的45°孔道内,分布在距中子源不同距离处,辐照结束后,采用HPGe探测器测量238 U圆片活化γ射线。实验结果与蒙特卡罗程序模拟计算结果进行了比较和分析。结果表明,238 U(n,2n)反应率实验结果与模拟计算值较吻合,238 U(n,2n)反应率随球体半径r的增加,近似服从e-ar/r2分布规律。     

Method for precision measurement of the masses of short-lived nuclides with application of stable isobars

The distinction of the suggested scheme of the experiment is the application of the isobaric doublets method in determining the masses of short-lived nuclides by means of a high-resolution mass spectrometer. This method enables us to considerably decrease the systematic errors of the measurement results in comparison with the previously applied isotopic reference method.

The work was carried out on an experimental complex, the main component of which was a prism mass spectrometer (PMS) with resolution R = 3 × 10⁴, installed on-line with a mass separator and a synchrocyclotron of the LNPI of the Academy of Sciences of Russia. On this installation we measured the masses of the isotopes ^91–97Rb. Samples containing stable reference isotopes ^91,92,94,96Zr, ⁹³Nb or ^95,97Mo were introduced into the ionizer of the PMS for calibration of the mass scale, that was ensured by their ionization simultaneously with radioactive isotopes of rubidium coming from the mass separator. The values of the masses of the stable isotopes used as reference in the suggested method are known with a very high precision (ΔM/M 3×10⁻⁸), therefore they made a negligible contribution to our experimental errors.     

Measurement of β-ray maximum energy with an HPGe detector by means of β-annihilation photons coincidence

A β⁺-ray detection system free from summation of annihilation photons has been constructed for the determination of Q_EC-values. It consists of an HPGe β-ray detector and two pairs of BaF₂ scintillation detectors for annihilation photons. A Q_EC-value of 4.83(4) MeV is obtained for ¹²⁶Cs separated with the JAERI on-line isotope separator.     

Cross-section measurements for the Mo (n,p) Nb reaction at the neutron energies from 13.6 to 14.9 MeV

Cross sections were measured at neutron energies from 13.6 to 14.9 MeV for the reaction ⁹⁷Mo(n,p)^97mNb leading to isomer of Niobium-97 isotope. The production of relatively short-lived isomer activity and the spectra accumulation have been carried out by cyclic activation method. Corrections were made for the effects of gamma ray attenuation, random coincidence summing (pulse pileup), dead time, neutron flux fluctuations and scattered low energy neutron contribution. Statistical model calculations for which the pre-equilibrium emission effects are taken into consideration were also performed for the investigated reaction between 13.0 and 15.0 MeV neutron energy range.     

秦山第二核电厂阀门类设备弱贯穿辐射调查

在秦山第二核电厂8次大修期间,对反应堆冷却剂系统(RCP)、余热排出系统(RRA)、化学和容积控制系统(RCV)、反应堆换料水池和乏燃料水池冷却和处理系统(PTR)4个系统主要阀门的辐射源项和弱贯穿辐射进行了监测。测得RCV系统阀门沉积的放射性核素主要是^110mAg,RCP、RRA和PTR系统阀门内沉积的主要是⁵⁸Co、⁶⁰Co、⁵¹Cr、⁹⁵Nb、⁹⁵Zr等放射性核素,伴随的β射线能量主要在500 keV范围内。测量给出了4类阀门的公式′(0.07)和H·′(3)值,测得H·′(3) /H·^*(10)值在1.24左右,H·′(0.07) /H·^*(10)值在14左右。结合测量结果,给出了部分阀门需要对检修人员开展眼晶体剂量和皮肤剂量监测的建议。     

Prediction of (n,3He) cross-sections around 14 MeV

A comprehensive review of the neutron-induced cross-sections for (n,³He) reactions has been made for the interval of 14⩽Z⩽84 around 14 MeV neutron energy. For practical purposes, an empirical expression has been found by using the experimental (n,³ He) cross section values as a function of (N-Z) and (E_n-E_th) where (N-Z) is the neutron excess of the target nucleus, E_n and E_th are the incident neutron energy and the (n,³He) threshold energy, respectively. The derived empirical relation gives a good fit with the experimental values     

Measurements and model calculations of activation cross sections for 232Th(n,2n)231Th reaction between 13.57 and 14.83 MeV neutrons

In this study, activation cross sections were measured for the reaction of ²³²Th(n,2n)²³¹Th (T_1/2 = 25.5 h) by using neutron activation technique at six different neutron energies from 13.57 and 14.83 MeV. Neutrons were produced via the ³H(²H,n)⁴He reaction using SAMES T-400 neutron generator. Irradiated and activated high purity Thorium foils were measured by a high-resolution γ-ray spectrometer with a high-purity Germanium (HpGe) detector. In cross section measurements, the corrections were made for the effects of γ-ray self-absorption in the foils, dead-time, coincidence summing, fluctuation of neutron flux, low energy neutrons. For this reaction, statistical model calculation, which the pre-equilibrium emission effects were taken into consideration, were also performed between 13.57 and 14.83 MeV energy range. The cross sections were compared with previous works in literature, with model calculation results, and with evaluation data bases (ENDF/B-VII, ENDF/B-VI, JEFF-3.1, JENDL-4.0, JENDL-3.3, and ROSFOND-2010).     

240Am半衰期的测量

²⁴⁰Am的半衰期对准确测量²⁴¹Am(n,2n)²⁴⁰Am反应截面具有重要作用,当前评价的数据50.8(3) h是对²⁴⁰Am的987.8 keV γ射线用Ge(Li)探测器跟踪测量6 d的结果,测量时间不到3个半衰期,使得测量结果的不确定度偏大。本文利用Geant4模拟软件建立了阱型HPGe探测器的测量模型,模拟计算了不同Pb吸收厚度下²⁴⁰Am高能γ射线的探测效率,确定使用阱型HPGe探测器配合吸收X射线和低能γ射线的Pb吸收体可有效提高²⁴⁰Am高能γ射线的探测效率。根据Geant4模拟计算的结果,Pb吸收体厚度为1 mm时,对²⁴⁰Am的888.8 keV和987.8 keV两条特征γ射线的探测效率分别为14.1%和13.3%。在中国原子能科学研究院的HI-13串列加速器上通过²⁴²Pu(p,3n)反应生产了²⁴⁰Am,制备了约700 Bq的²⁴⁰Am测量源,用上述方法跟踪测量²⁴⁰Am的888.8 keV和987.8 keV两条特征γ射线的强度,时间超过18 d,用最小二乘法拟合得到其半衰期为50.79(5) h,结果与评价结果一致,但减小了不确定度。     

液体闪烁谱仪测量~(90)Y半衰期及其不确定度分析

~(90) Y是一种非常重要的治疗用放射性核素,准确测定~(90) Y的半衰期对理论研究和实际应用均具有重要意义。目前对~(90 )Y半衰期的测定已有大量文献报道,但是测定结果不确定度较大,从63.46h到64.6h不等。本工作利用液体闪烁谱仪对HDEHP-甲苯直接萃取的纯~(90) Y半衰期进行测定,并给出了结果的不确定度。~(90) Y半衰期的测定结果为64.05h,不确定度为0.017%。     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

Cobalt cross sections for 14 MeV neutrons

14 MeV cross sections for the reactions ⁵⁹Co(n,2n) ^58m+gCo, ⁵⁹Co(n,p) ⁵⁹Fe and ⁵⁹Co(n,) ⁵⁶Mn were measured relative to the ⁵⁶Fe(n,p) ⁵⁶Mn reaction employing the activation technique. Accuracies of about 1% were achieved for the (n,) reaction and 2% for the others. The isomeric cross section ratio was measured for the ⁵⁹Co(n,2n) reactions.

Nuclear reactions – ⁵⁹Co(n,2n) ^58m+gCo, ⁵⁹Co(n,p) ⁵⁹Fe, ⁵⁹Co(n,) ⁵⁶Mn, E_n = 14.3, 14.7 MeV: measured activation cross sections relative to ⁵⁶Fe(n,p) ⁵⁶Mn.

⁵⁹Co(n,2n) ^58m,gCo, E_n = 14.3 MeV; measured isomeric cross section ratio. Natural targets. Ge, NaI and 4πβ detectors.     

Fusion-driven transmutation of selected long-lived fission products

The long-term radiological burden associated with nuclear power production is usually attributed to long-lived fission products (LLFP). Their lifetime and large equilibrium mass and hence radioactivity accumulated in the course of fission energy generation make their storage a rather formidable task to solve. Therefore the idea of artificial incineration of LLFP through their transmutation has been quite naturally incorporated into the concept of self-consistent nuclear energy system (SCNES) based primarily on fast breeder reactor technologies. However it is now acknowledged that neutron environment of fission facilities including fast breeder reactors does not seem most appropriate for LLFP transmutation. The issue has been then extensively developed within the framework of multi-component self-consistent nuclear energy system (MC-SCNES). Neutrons of specific quality required for LLFP transmutation are proposed there to be of non-fission origin. Given neutron excess available and neutron quality, a fusion neutron source (FNS) is appearing as the candidate No. 1 to consider for LLFP transmutation. Research on LLFP transmutation by means of FNS has very long history and has received an additional boost during the decade passed. In the present study, potential of thermal flux blanket of FNS is exemplified by transmutation of ⁹³Zr and ¹²⁶Sn, the most difficult LLFP to transmute. It is shown that transmutation of ⁹³Zr is effective even with a rather modest neutron loading of 1 MWt·m⁻², typical for ITER project. Transmutation of ¹²⁶Sn, however, requires neutron loading of as high as 3 MWt·m⁻² for DD fusion and is quite unattractive for DT fusion. In the latter case, transmutation through the threshold (n,2n) reaction may be preferable.     

Improved performances of the fast reactor calculational system ERANOS-ERALIB1 due to improved a priori nuclear data and consideration of additional specific integral data

A single consistent scheme of calculational methods and nuclear data called ERANOS-ERALIB1 was produced in 1996 to calculate fast reactor neutronic parameters. It represents a significant improvement on previous schemes such as CARNAVAL-IV, PROPANE and VASCO, each of which were required in order to calculate one specific application. The nuclear data library ERALIB1 has been obtained by a consistent statistical adjustment based on 355 integral data from 71 different systems. The performance of ERALIB1 is excellent, as demonstrated during its validation for which all the k_eff SUPER-PHENIX data were reproduced to within 70 pcm.

The only restriction on this satisfactory performance is related to the rather poor prediction of the sodium void reactivity effect. This was due to very bad nuclear data for ²³Na, and the unsatisfactory methods used to calculate the sensitivity coefficients for the sodium void reactivity variation Δρ_Na. To improve the performance relative to this point and to enlarge the domain of validation several actions have been undertaken:

•a revision of the formalism and algorithms used to calculate the derivatives of Δρ_Na to the sodium cross section data,

•a significant enlargement of the integral data base related to this aspect of the sodium void effect. Compared to the initial data base established in support of ERALIB1, several additional (18) sodium void configurations corresponding to voids of different volumes at different core locations have been studied.

In order to broaden the range of application of the improved library, which will be called ERALIB1.A, significant effort has been devoted to additional configurations which have firstly been evaluated, and then if judged suitable, included in the adjustment process. They are related to two specifically targeted experimental programmes:

•a study of neutron deep penetration. Several configurations of the JANUS experimental programme (shielding constructed of separate steel, iron and sodium plates) have been analysed. With this complementary information ERALIB 1A becomes applicable for accurate predictions of shielding configurations,

•a study of steel reflectors for a fast reactor of the SUPER-PHENIX type. The measurements performed in the MASURCA (the CIRANO experimental programme) and FCA facilities include spectral indices (F25(r)/F25(0),...) at different positions in the reflector. As a consequence of these measurements, important information has been obtained for additional “secondary” structural material isotopes, such as ⁵⁷Fe, ⁶⁰Ni and ⁵³Cr.

Significant effort has also been devoted to the analysis of 29 β_eff experiments. The result of this is an improvement of the uncertainty on ν_d(E) which guarantees a prediction of β_eff with the required accuracy (3% for critical configurations, and 5% for power reactors).

The consistent statistical adjustment method by Gandini et al. (1973) has been completed. Rigorous criteria have been introduced to identify any data which are suspect and/or inconsistent in the integral data base. These data may introduce additional bias in the adjusted library, and for that reason they must be discarded before adjustment.     

14.8 MeV中子诱发78Kr(n,2n)77Kr反应截面的测量

快中子诱发(n,2n)反应截面的测量在核反应机制研究和核技术应用等方面有着广泛的应用价值。本文在中国原子能科学研究院的高压倍加器上,基于活化法实验测量了78Kr(n,2n)77Kr在148 MeV能点的反应截面。样品靶为高纯78Kr气体样品,用十万分之一天平称重得到78Kr的质量,将两片高纯93Nb薄片分别固定在样品靶两侧以监测中子注量率。利用T(d,n)4He反应产生148 MeV中子,轰击距中子源约10 cm的样品靶大于4 h后,用准确刻度过效率的HPGe探测器测量活化产物 77Kr和92Nbm的活度。利用蒙特卡罗程序计算中子注量率修正、样品自吸收修正、样品几何修正等因子,得到了78Kr(n,2n)77Kr的反应截面,并将结果与文献值和评价数据库进行了比较。研究结果有助于提高78Kr(n, 2n)77Kr反应截面测量和评价的水平。     

keV能区中子注量率测量初级标准装置研制

由于大多数中子探测器在keV能区处于共振能区,截面数据无法精确已知,无法作为中子注量率测量初级标准装置。⁶Li在keV能区的截面精确已知,本文以此设计了一种包含⁶LiF转换体的硅半导体探测器系统（LiF-SSD探测器）,建立了keV能区中子参考辐射场中子注量率测量初级标准装置。利用SRIM和TRIM程序估算了LiF-SSD探测器的计数率;基于Geant4开发了探测器响应计算程序,给出了中子能量为27.4 keV的探测器脉冲幅度计算谱;利用⁴⁵Sc(p,n)-⁴⁵Ti反应,测量了27.4 keV中子能量点的探测器脉冲幅度谱,测量结果与理论计算谱符合较好。     

~(176)Hf(n,2n)~(175)Hf反应截面测量

采用相对测量技术,以活化法对13.4～14.8MeV范围内的176Hf(n,2n)175Hf反应截面进行了测量。样品固定在距离D-T中子源20cm处的圆环的不同位置上进行中子辐照,采用93Nb(n,2n)92Nbm作为监测反应,活化产物采用高纯锗探测器进行了测量,所得14MeV附近的176Hf(n,2n)175Hf反应截面实验值为(2100±85)mb,对实验结果与公开文献值和ENDF/B6.8评价库数据进行了比对。