CARR堆冷中子瞬发伽玛活化分析系统及实验研究

利用中国先进研究堆（CARR）在国内首次开展了冷中子瞬发伽玛活化分析（CNPGAA）实验,采用定制加长的电制冷高纯锗（HPGe）探测器和先进的数字多道谱仪DSPEC®-502进行测量,获得了NH₄Cl样品中元素冷中子瞬发伽玛谱和本底谱等数据,同时利用伽玛放射源¹⁵²Eu、¹³⁷Cs、⁶⁰Co以及NH₄Cl产生的瞬发伽玛射线对探测器在宽能区0.1~8 MeV进行能量刻度。为降低环境辐射本底,HPGe探测器外围采用环形锗酸铋（BGO）康普顿谱仪,10 cm铅以及含⁶Li和¹⁰B材料对中子束流准直屏蔽。此外,利用金片活化法测量了CARR堆运行功率为15 MW时有无冷源情况下冷中子导管B(CNGB)末端1 m处的中子注量率,结果显示有冷源时中子注量率可提高一个量级。     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

A digital spectrometer approach to obtaining multiple time-resolved gamma-ray spectra for pulsed spectroscopy

Neutron-induced gamma-ray emission and its detection using a pulsed neutron generator system is an established analytical technique for quantitative multi-element analysis. Traditional gamma-ray spectrometers used for this type of analysis are normally operated either in coincidence mode - for counting prompt gamma-rays following inelastic neutron scattering (INS) events when the neutron generator is ON, or in anti-coincidence mode - for counting prompt gamma-rays from thermal neutron capture (TNC) processes when the neutron generator is OFF. We have developed a digital gamma-ray spectrometer for concurrently measuring both the INS and TNC gamma-rays using a 14 MeV pulsed neutron generator. The spectrometer separates the gamma-ray counts into two independent spectra together with two separate sets of counting statistics based on the external gate level. Because the TNC gamma-ray yields are time dependent, additional accuracy in analyzing the data can be obtained by acquiring multiple time-resolved gamma-ray spectra at finer time intervals than simply ON or OFF. For that purpose we are developing a multi-gating system that will allow gamma-ray spectra to be acquired concurrently in real time with up to 16 time slots. The conceptual system design is presented, especially focusing on considerations for tracking counting statistics in multiple time slots and on the placement of pulse heights into multiple spectra in real time.     

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

Effect of the VVR-SM neutron spectrum on the radioactivity and color of natural topazes

The neutron flux density from 0.025 eV to 12 MeV has been measured experimentally in all channels of the VVR-SM core by the activation method using threshold monitors (Au, Ni, Fe. Ti, Mg, Y). Comparing with a calculation of the neutron flux density at different energy using the IRT-2D computer code showed agreement to within 5%. The distribution of the neutron fluxes and spectra in the core, which is of practical utility for radiation technologies, was obtained. A series of irradiations has been conducted and experimental dependences of the irradiation time on the channel position in the core as well as on the size of the stones for obtaining a standard light blue and dark blue color have been obtained. The irradiation conditions making it possible to lower the induced radioactivity of the minerals three-fold as a result of increasing the ratio of the fast to thermal neutron fluxes are found.     

Characterization of an Am-Be PGNAA set-up developed for in situ liquid analysis: Application to domestic waste water and industrial liquid effluents analysis

A prompt gamma neutron activation analysis (PGNAA) set-up with an Am-Be source developed for in situ analysis of liquid samples is described. The linearity of its response was tested for chlorine and cadmium dissolved in water. Prompt gamma efficiency of the system has been determined experimentally using prompt gamma of chlorine dissolved in water and detection limits for different elements have been derived for domestic waste water. A methodology to analyze any kind of liquid is then proposed. This methodology consists mainly on using standards with water as bulk or in the case of absolute method, to use gamma efficiency determined with prompt gammas emitted by chlorine dissolved in water. To take into account the thermal neutron flux variations inside the samples, flux monitoring was carried out using a He-3 neutron detector placed at the external sample container surface. Finally, to correct for the differences in gamma attenuation, average gamma attenuations factors were calculated using MCNP5 code. This method was then checked successfully by determining cadmium in industrial phosphoric acid and our result was in good agreement with that obtained with inductively coupled plasma (ICP) method.     

Measurements of the neutron yield and the neutron energy distribution from the 9Be(d,n)10B reaction on a thick Be target at an incident deuteron energy of 20.2 MeV

A fast neutron irradiation facility has been set up at the SARA cyclotron located in Grenoble. This facility provides the possibility to carry out fast neutron irradiation tests at both cryogenic and ambient temperatures. Neutrons are produced by stopping a 20.2 MeV deuteron beam in a 3 mm thick beryllium target. The angular distribution of the neutron flux and the energy spectra from 0° to 40° with respect to the deuteron beam axis were measured. A neutron fluence in the range of 10¹⁴ cm⁻² is available per day, with a small gamma contamination and a small thermal neutron flux.     

The Monte Carlo code CEARCPG for coincidence prompt gamma-ray neutron activation analysis

Prompt gamma-ray neutron activation analysis (PGNAA) is widely used to determine the elemental composition of bulk samples. The detection sensitivities of PGNAA are often restricted by the inherent poor signal-to-noise ratio (SNR). There are many sources of noise (background) including the natural background, neutron activation of the detector, gamma-rays associated with the neutron source and prompt gamma-rays from the structural materials of the analyzer. Results of the prompt gamma-ray coincidence technique show that it could greatly improve the SNR by removing almost all of the background interferences. The first specific Monte Carlo code (CEARCPG) for coincidence PGNAA has been developed at the Center for Engineering Application of Radioisotopes (CEAR) to explore the capabilities of this technique. Benchmark bulk sample experiments have been performed with coal, sulfur, and mercury samples and indicate that the code is accurate and will be very useful in the design of coincidence PGNAA devices.     

Analysis of oxygen content in steel by means of 14 MeV neutrons

A sealed-tube type 14 MeV neutron generator with maximum neutron output of 10¹¹ n/sec, incorporating a pneumatic sample transfer system of single-tube type and with a single rotation of the sample during neutron irradiation, is used to develop a method for determining oxygen content in steel in the ppm range with the best precision, and at the same time, to make the process suitable for routine work in industrial applications. The pneumatic sample transfer system is made to incline at an angle of about 20° towards the horizontal at the irradiation station. Together with a constant pressure gas reservoir for providing a constant optimum gas pressure in the transfer tube, the system gives a result of nearly perfect reproducibility in the operation.A pulse shape analyser system incorporating an organic scintillation detector is used for monitoring neutron flux level during the neutron irradiation of the sample. The percentage standard deviation of the neutron counts by the present monitoring system ranges from 0.9 to 2.7% with 0.5% as the percentage statistical deviation alone.Polyethylene, of oxygen content 163 ppm determined by comparison with lucite, is used as the steel sample carrier. A 3 × 3 in. NaI(Tl) crystal is used with a single channel analyser to count the 6.1 and 7.1 MeV gamma rays emitted from ¹⁶N as a result of the reaction ¹⁶O(n, p)¹⁶N. An optimum combination for the time of irradiation, delay and counting of the induced activity; of 30, 0.1 and 30 sec, respectively, is chosen in the present experiment. Thus, for a 100 g steel sample with an oxygen concentration of 170 ppm, the percentage standard deviation is about 4.4% which is, in fact, the counting statistic itself, resulting from a neutron flux level of 1.3 × 10⁸ n/sec cm² at the sample. As the present activation analysis makes use of the comparison method, a steel-mylar standard made of layered steel and mylar discs is prepared and a calibration curve constructed. A method of correcting the oxygen contribution in the polyethylene sample carrier is devised and the content of oxygen in the steel standard is determined.A survey of neutron flux distribution is also attempted and it is found that nearly symmetrical distribution of the flux, about the centre of the sample carrier which is placed with its axis in parallel to the plane of the disc-shaped target of the neutron-generating tube, is far from being flat.     

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。     

A thin film 10B sample for measuring the atom number

In order to measure differential cross sections of the 10B(n,α)7Li reaction induced by MeV neutrons using the forward-backward coincidence method, a thin film 10B sample was designed and the 10B atom number was determined with a reference 10B film sample. Alpha counts of the 10B(nth,α)7Li reaction from the 10B thin film and the reference sample were measured using a gridded ionization chamber and thermal neutrons, which were moderated and thermalized by paraffin from fast neutrons produced in D(d,n)3He reaction on a 4.5 MV Van de Graaff. The neutron flux was normalized by measuring the fission yield of a small 238U fission chamber.     

Thermal, structural and mechanical properties of neutron irradiated Bayfol nuclear track detector

Samples from sheets of the polymeric material Bayfol have been exposed to neutrons of incident energy in the range 0.8-19.2 MeV. The resultant effect of neutron irradiation on the thermal properties of Bayfol has been investigated using thermo-gravimetric analysis. The onset temperature of decomposition and activation energy of thermal decomposition were calculated. The variation of transition temperatures with neutron energy has been determined using differential thermal analysis. The results indicate Bayfol thermograms characterized by the appearance of an endothermic peak due to melting. Melting temperature was found to be dependent on the neutron energy. Structural property studies using infrared spectroscopy were performed and results indicated that scission takes place at the carbonate site with the formation of a hydroxyl group. Mechanical properties were studied and it is shown that, at the fluence range 0-4.4 MeV, the standard chains and a great number of chain ends weaken and the material may become softer.     

瞬发伽玛活化分析中3种探测器性能比较

利用中国先进研究堆(CARR)热中子束流孔道首次开展了瞬发伽玛中子活化分析(PGNAA)实验。对NH4Cl、Si、Fe、Al等4种样品进行了辐照,同时采用HPGe、LaBr_3、BGO 3种探测器对样品进行实时测量,在瞬发伽玛射线的能量为0.002~10 MeV范围内研究了3种探测器在宽能区的能量线性、能量分辨率、探测效率等性能。     

Experimental optimization of a landmine detection facility using PGNAA method

The optimum moderator geometry increases the performance of prompt gamma neutron activation analysis (PGNAA) method considerably. In this work an 241Am-Be source was used in the moderator geometry for detecting buried landmines by PGNAA method. Experiments were done to find the best moderator geometry for the moderated 241 Am-Be source, by replacing the mine with a neutron detector and counting the thermal neutron flux. The flux of thermal neutrons at the place of mine was used as a determining factor to introduce the best moderator geometry.     

Proton simulation of displacement effects induced in metals by 14 mev neutrons

In designing a D-T fusion reactor, one must know the effect of a high flux of 14 MeV neutrons on structural materials. Available laboratory sources of 14 MeV neutrons are not intense enough to expose samples to the expected flux. Bombardment with other particles is one way of simulating the anticipated neutron environment. The energy spectrum of atoms recoiling from collisions with bombarding particles can be calculated from elastic-scattering and nonelastic-reaction data for the incident species. This analysis shows that 16 MeV protons closely simulate the displacement effects caused by 14 MeV neutrons. In niobium the average atom recoiling from a 14 MeV neutron interaction has 65 keV of damage energy. The mean damage energy deposited per cm³ of niobium by a fluence of one 14 MeV neutron per cm² is 14 keV. The equivalent quantity for 16 MeV protons incident on niobium is 33 keV.     

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.     

Problems of lead nuclear data in fusion blanket design

In an irradiation experiment using a LiAl/Pb assembly, we found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10–16 MeV region by around 30% and that in the 0.5–5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5–10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: the cross-sections of the (n,2n) reaction in JENDL-3.3 for lead isotopes are too large and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for ²⁰⁸Pb in JENDL-3.3 reproduce previous experimental double-differential cross-section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5–10 MeV with JEFF-3.1 is still unclear.     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Neutron fluence rate measurements in a PGNAA 208-liter drum assay system using silicon carbide detectors

Pulsed prompt gamma neutron activation analysis (PGNAA) is being implemented for the nondestructive assay (NDA) of mercury, cadmium and lead in containers of radioactive waste. A PGNAA prototype system capable of assaying 208-liter (55-gallon) drums has already been built and demonstrated. As part of the evaluation of this system, the thermal neutron fluence rate distribution in a drum containing a combustible waste surrogate was measured during PGNAA runs using a silicon carbide neutron detector. The fast charge-collection time of this detector type enabled the investigation of the neutron kinetics at various locations within the matrix during and between pulses of the system’s 14-MeV neutron source. As expected, the response of a SiC detector equipped with a lithium-6 fluoride layer is dominated by thermal neutron-induced events between pulses. The measurement results showed that the thermal neutron fluence rate is relatively uniform over a radial depth of several centimeters in the matrix region that contributes a significant fraction of the prompt gamma radiation incident on the system’s photon detector.     

LBS—9型裂变室的堆内辐照性能

本文介绍了用于宽量程监测装置的裂变室的辐照性能,给出了它在脉冲计数工作时的性能与涂铀量、瞬时辐照通量和累积辐照通量的关系。大量实验证实,LBS-9型裂变室能可靠地用于宽量程监测装置。