TIMS测定辐照后Gd同位素比的方法研究

建立了热电离质谱法(TIMS)测量天然及辐照后氧化钆同位素丰度比的检测方法。天然氧化钆制备成靶件,放入高通量工程试验堆(HFETR)预定孔道接受中子辐照,辐照时间共计91.3 h,反应堆功率为80 MW,辐照孔道中子注量率约为2×10~(14) n·cm~(-2)·s~(-1)。辐照后靶件经切割、转运、溶解、制样与测量等过程,完成了对可燃中子毒物钆的辐照与测量,并对测量值进行了修正。数据结果表明,热电离质谱法对Gd辐照后检验的分析数据准确可靠,辐照后样品的后处理方法合理,钆的各同位素的丰度变化值前后吻合,且与中子吸收截面大小密切相关。本方法可用于铀钆混合燃料芯块和卸料元件组件中可燃毒物钆的同位素分析。获得的实测数据可反向用于理论计算修正,以期获得更优的反应堆堆芯设计方案。     

富集同位素用于长循环堆芯可燃毒物的中子学分析

长循环堆芯候选弥散型可燃毒物Sm_2O_3、HfO_2、Dy_2O_3、Eu_2O_3及Er_2O_3在寿期末具有一定残留,影响燃料经济性。从反应性补偿角度对寿期初引入负反应性、寿期内消耗速率和寿期末反应性残留3方面研究分析了富集同位素~(149)Sm、~(177)Hf、~(164)Dy、~(151)Eu、~(167)Er用于可燃毒物的可行性。研究表明采用富集同位素作为可燃毒物有一定优势:同位素富集后,弥散型可燃毒物装量下降,降低了对燃料芯块性能的影响;寿期内反应性曲线更加平缓,有利于反应性控制;寿期末可燃毒物残留有较大改善,提高堆芯燃料经济性。     

中国实验快堆生产同位素~(60)Co的优化设计研究

中国实验快堆(CEFR)是我国建设的第一座金属钠冷却快中子反应堆,其特点是比功率高、中子注量率高。由于CEFR能谱硬、中子泄漏多,本工作对利用这部分泄漏中子在快堆反射层生产~(60)Co同位素进行了可行性研究,并通过引入慢化剂设计适合的几何结构、选择合理的辐照位置和辐照时间等一系列措施对同位素生产进行优化,以提高~(60)Co比活度。优化中采用MCNP和ORIGEN2程序作为计算软件。计算结果表明,通过优化设计,~(60)Co比活度有显著提高,说明利用快堆反射层的泄漏中子生产~(60)Co是可行的。     

铁同位素窄电离带质谱分析

本工作采用0.4 mm宽的窄电离带,减小了由于电离带表面高温热辐射引起的本底干扰。在电离带和样品带的加热电流分别为4.0 A和1.8 A条件下,用5μg氧化铁样品可以获得10~(-13)A稳定的离子束流近1小时。用这种方法测定了电磁分离器生产的各种浓缩铁同位素产品的丰度,测定精度为0.1%。     

FCM燃料压水堆弥散可燃毒物中子学分析

为将全陶瓷微胶囊封装(FCM)燃料应用于小型压水堆,对FCM燃料组件开展了可燃毒物中子学设计与分析。通过寿期初引入负反应性、寿期内消耗速率和寿期末残留3个方面,对弥散在SiC基体中的弥散型可燃毒物Gd_2O_3、Er_2O_3、Sm_2O_3、Eu_2O_3、Dy_2O_3及HfO_2进行评价。FCM燃料中TRISO颗粒核芯直径达800μm,燃料颗粒自屏效应强烈,在RMC程序中引入随机介质计算功能,对FCM燃料进行随机几何建模,保证了反应性计算精度。分析表明:Er_2O_3可作为FCM燃料堆芯的候选可燃毒物,Gd_2O_3和Eu_2O_3需结合堆芯开展进一步研究,Sm_2O_3、Dy_2O_3及HfO_2的反应性惩罚过大,不适合作为FCM燃料可燃毒物。     

动力堆辐照元件中燃耗测定和~(149)Sm,~(150)Sm含量,~(154)Eu/~(155)Eu,~(154)Eu/~(152)Eu比值与燃耗的关系

本文用同位素稀释质谱法,以~(148)Nd为燃耗监测体对某动力堆元件的燃耗进行了测定。还测定了裂变产物中的高中子毒物~(149)Sm的含量。对~(150)Sm的含量测定结果表明,它能反映出核燃料燃烧的程度,为一直线关系。γ谱法测得的~(154)Eu/~(155)Eu比值和燃耗呈曲线关系。     

用放射化学中子活化方法研究陨石中的Os和Ru同位素丰度

选择南丹、邕宁铁陨石经2mol/l H_2SO_4处理后的残渣,和吉林、洮南石陨石的金属相部分,以Os、Ru试剂(地球同位素丰度)为标准,用放射化学中子活化方法测定~(190)Os/~(184)Os及~(96)Ru/~(102)Ru的比值。实验全面考虑了在照射、分离、测量过程中各种因素对Os和Ru同位素丰度测定的影响。对~(190)Os/~(184)Os的测量统计误差可控制在1％以内。实验结果表明,这4种陨石的~(190)Os/~(184)Os和~(96)Ru/~(102)Ru的比值与地球标准相比,不存在统计上有意义的同位素丰度比异常。     

天然钐和浓缩钐同位素丰度比质谱测定

裂变产物~(149)Sm热中子吸收截面大,是反应堆中妨碍提高反应性的可燃毒物。用MATCH5质谱计,测定了天然氧化钐和浓缩钐同位素丰度比和百分原子浓度,~(149)Sm同位素丰度比测定精度在0.4％左右,为同位素稀释质谱法准确测定~(149)Sm绝对浓度奠定了基础。     

~(169)Yb两种扫描剂的制备

本文介绍~(169)Yb(镱)的两种扫描剂~(169)Yb-DTPA和柠檬酸~(169)Yb的制备和分析鉴定。采用国产光谱纯Yb_2O_3(天然丰度)作靶材料,在反应堆中经高通量中子照射后,进行化学处理,最后产品的放射性核纯和放化纯度均大于99%,放射性比度2～5毫居里/毫克镱,放射性浓度≥5毫居里/毫升。确定了DTPA与Yb~(3 )螯合时的分子比是1:1,产生~(169)Yb的有效反应截面实验值约2600靶。讨论了这两种制剂在诊断上的适用性。     

硼硅玻璃可燃毒物的辐照试验

描述了硼硅玻璃可燃毒物的辐照考验及辐照后性能检验,其中包括辐照后试样的外观、腐蚀检查、尺寸及密度测量、氦气释放量和~(10)B燃耗测定,对玻璃腐蚀和辐照密实现象进行了讨论。通过两种规格的样品进行不同中子注量的辐照考验,证明国产GG-17硼硅玻璃取代~(10)B不锈钢用作秦山核电站可燃毒物是安全、经济、合理的。     

Isotopic Enrichment and Thermal Neutron Cross Section Measurement of Erbium-167

The thermal neutron cross section of Er167 has been obtained by measuring the relative number of Erl68 atoms formed by radiative neutron capture, when isotopically pure Er167 samples were exposed to a highly-thermalized neutron flux. A four-stage mass spectrometer was used to provide a 106/1 isotopic enrichment of Er167. It was also employed it measure the postirradiation Er168/Er167 isotopic ratios of the samples, which were placed in the D2O moderated Argonne CP-5 reactor. The integrated neutron flux, and effective neutron temperature in this reactor were also monitored mass spectrometrically, by observing isotopic ratio changes in nuclides whose cross section as a function of temperature are known. Two rare earth isotopes Sm149 and Gd157, were selected as the "non-l/v" absorbers that yielded this data. The effective neutron cross section of Er167 was found to be 740 ± 21 barns, at an effective neutron temperature of 42.8 + 2.5°C, with an integrated neutron flux of 3.37 ± 0.08 × 1019 neutron/cm2. The thermal (2200 m/sec) neutron cross section of Er167 was then calculated to be 699 ± 20 barns.     

Dissolution and analysis of erbium doped cermet fuels

Cermets are suggested as new kind of nuclear fuel to reduce global costs. They need high enriched fuel and thus use of burnable poison. Special pellets were developed and irradiated to test such concepts. Some pellets consist of a cermet fuel. With an improved fuel thermal conductivity (by using metal matrix), lower temperatures than standard fuel are obtained. Some pellets were made of cermet and erbium in small quantity. Studies on erbium were launched to determine the influence of this neutron poison. Standard dissolutions (HNO₃, HF) on cermet (Mo-UO₂) and on erbium doped cermet show a large amount of insoluble matter. Tests have been carried out in order to establish a procedure for a complete dissolution of active pellets. Consequently, an optimal process was defined. Irradiated pellets from experimental reactor SILOE will be dissolved. Analytical chemistry studies were undertaken. Thermal Ionization Mass Spectrometry (TIMS) and Glow Discharge Mass Spectrometry (GDMS) have been applied. The U and Er isotopic composition has been determined on different samples.     

Design concept of a small nuclear reactor for large-diameter NTD-Si using a conventional PWR full-length fuel assembly

A design concept for a small nuclear reactor dedicated to large-diameter neutron transmutation doping silicon (NTD-Si) is proposed. Conventional PWR (Pressurized Water Reactor) full-length fuel assembly is used to assure stable and reliable supply of fuel. Criticality, neutron transportation, and core burn-up calculations are performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the proposed reactor can be critical over 18 years, and excess reactivity can be suppressed by a combination of Gd₂O₃ burnable poison and soluble boron. Preliminary steady-state single-channel thermal hydraulic analysis showed that heat removal from core is possible under 1 atm operating pressure. Si ingots up to 30 cm in diameter can be irradiated in the reactor irradiation channels, and the uniform irradiation condition can be achieved for a large-diameter Si ingot.     

Evolution of the neutron sensor characteristics in the RBMK-1500 reactor neutron flux

Variation of characteristics of the RBMK-1500 reactor radial neutron flux sensors with the HfO₂ emitter during long-term maintenance was investigated. The influence of nuclear fuel enrichment and burnable erbium admixtures on the energy neutron spectrum, neutron absorption, and hafnium isotopic composition variation was considered. The dependences of corrective factors of the neutron sensor signal on the nuclear fuel burnup depth and the integral current accumulated by the sensor for different enrichment nuclear fuel are presented in the work. The experimental verification of the calculated dependence of the sensor corrective factor on the accumulated integral current was performed.     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

惰性基质燃料元件Doppler系数分析

本文采用中子输运程序MCNP,基于ENDF/B-Ⅶ-1核数据库,对几种典型惰性基质燃料（IMF）的Doppler系数进行了计算,并通过理论分析给出了各核素对Doppler系数贡献的表达式。结果表明：在相同惰性基质条件下,武器级Pu燃料的Doppler系数的绝对值小于反应堆级Pu燃料的;在惰性基质中添加²³²Th可使Doppler系数更负,且可使IMF获得与低浓UO₂燃料相近的Doppler系数;硼可燃毒物对Doppler系数的贡献为正效应,而铒可燃毒物则可进一步增强负Doppler系数,有利于反应堆的固有安全性。     

Electrical characteristics of neutron irradiation induced defects in n-GaAs

Palladium Schottky barrier diodes (SBDs) on epitaxially grown n-GaAs were irradiated with neutrons from a reactor and a p(66)/Be (40) clinical source. From current-voltage (I–V) and capacitance-voltage (C–V) measurements it was found that neutron irradiation caused generation-recombination currents and resulted in a reduction in the free carrier concentrations of the epitaxial layers. A linear relation was found between the irradiation fluence, the free carrier removal and the reverse leakage current of neutron irradiated SBDs. Deep level transient spectroscopy (DLTS) indicated that five electron traps, Enl-En5, were introduced during neutron irradiation. These defects are shown to be responsible for the degradation of neutron irradiated SBDs.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.     

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

Measurement of the activation cross-sections of (n,np+d) reactions for lanthanide isotopes in the energy range between 13.4 and 14.9 MeV

The activation cross-sections for nine (n,np+d) reactions were measured by the activation method in the energy range between 13.4 and 14.9 MeV. The irradiated targets were lanthanide isotopes: ^146,148Nd, ¹⁵²Sm, ^{155, 158}Gd, ¹⁶⁴Dy, ¹⁷⁰Er, and ^174,176Yb. The cross-sections, except for ¹⁷⁰Er, were obtained for the first time. The D–T neutron source of the fusion neutronics source (FNS) at the Japan Atomic Energy Research Institute was used for irradiation. All cross-section values were determined relative to that of the ²⁷Al(n,α)²⁴Na reaction (ENDF/B-VI). To obtain reliable activation cross-sections, careful attention was paid to the corrections of the neutrons irradiation and induced activity measurements, as well as the effective neutron energy determination at the irradiation positions. To measure weak activities, a highly efficient measuring technique with a well-type HPGe detector was applied. The present results were compared with the comprehensive evaluated data in the JENDL-3.2, JENDL-Activation File, ENDF/B-VI and FENDL/A-2.0. Most of the evaluated data were overestimated or underestimated by more than 30%. Especially, there were the large underestimations for ¹⁶⁴Dy, ¹⁷⁰Er, ¹⁷⁴Yb and ¹⁷⁶Yb in FENDL/A-2.0.