Neutron Flux Ratio Research Articles

Reevaluation of neutron flux characterization parameters for NIST RT-2 facility

It has been difficult to characterize the thermal to epithermal neutron flux ratio (f) and the measure of the nonideal epithermal neutron flux distribution (α) for the RT-2 pneumatic rabbit facility at the NIST National Bureau of Standards Reactor (NBSR). In a previous paper, only cadmium-covered irradiations yielded physically reasonable parameters. New measurements were performed using chromium, manganese, cobalt, zinc, zirconium, molybdenum, antimony, gadolinium, lutetium, and gold. The neutron temperature (T n ) in RT-2 measured using bare lutetium and gold foils gave unphysical values. The bare foil methods for measuring f and α gave inconsistent results. The underlying reasons are demonstrated via MCNP simulation results for cumulative reaction rates of selected isotopes. To determine expected intervals for f, α and T, parametric methods were explored. Measured reaction rate probability per target atom (R p ) values for the listed elements were fitted to a modified Westcott curve using an iterative least-squares method to verify consistency of measurements and nuclear data. An advanced parametric approach using a detailed MCNP model of the NBSR was used to calculate neutron flux characterization parameters.

A new reflector structure for facility thermalizing D–T neutrons

A facility for thermalization of fast neutrons (14.2 MeV) emitted by compact deuterium–tritium (D–T) neutron generators (NGs) for thermal neutron activation analysis is proposed. Its final design is based on Monte Carlo calculations (MCNP5). To maximize the ratio between the thermal neutron flux and the total neutron flux and simultaneously to ensure the highest possible value of the thermal neutron flux at the output surface, the facility should consist of a two-layer reflector [tungsten (W)—the inner part, molybdenum—the outer part], a two-layer multiplier (W followed by lead), a moderator (polyethylene followed by magnesium fluoride) and a collimator (molybdenum and nickel near the output surface). For the D–T NG producing the maximum available neutron yield 1015 n s−1, the facility provides the thermal neutron flux 2.0 × 1011 n cm−2 s −1 and a slightly higher fast neutron flux 2.3 × 1011 n cm−2 s−1. To improve the ratio of the thermal neutron flux to the fast neutron flux (above 2.7) an addition of a silicon layer to the moderator and especially a proper adjustment and a threefold increase of the multiplier thickness is necessary.

Verification of k 0-NAA results at the LVR-15 reactor in Řež with the use of Au+Mo+Rb(+Zn) monitor set

Long-time experience in neutron flux moni- toring on irradiation in the LVR-15 research reactor in R y ez y proved that Au?Mn?Rb and Au?Mo?Rb(?Zn) monitor sets for short and long irradiation, respectively, are more suitable in our conditions than the most frequently used Au?Zr set. The advantages of the former monitor set have been described previously, in the present work we discuss the advantages of the latter monitor set for long irradiations in varying active core configurations of the LVR-15 reac- tor. The successful application of the Au?Mo?Rb(?Zn) monitor set has been verified by comparative determination of the neutron flux parameters a (epithermal flux distribu- tion parameter), f (thermal-to-epithermal neutron flux ratio), and Fc,Au (comparator factor) using this and the Au?Zr monitor set, and by analyses of certified reference materials, namely NIST SRMs 1547 Peach Leaves, 2711 Montana Soil, and 1633b Trace Elements in Coal Fly Ash.

Simulation and Design of a Simple and Easy-to-use Small-scale Neutron Source at Kyoto University

A simple and easy-to-use compact neutron source based on a low power level proton accelerator (proton energy 3.5 MeV and 0.35kW beam power) at Kyoto University was designed with the conception of low cost, compact size, high safety and intensive thermal neutron flux via Monte Carlo method with PHITS code. By utilizing (p, n) reactions in a beryllium target coupled to a polyethylene moderator and graphite reflector with a wing configuration, this facility is expected to produce time-averaged thermal neutron fluxes suitable for neutron scattering and development of instrumentation, and play a role in educating students in neutron science and performing research with neutrons. Borated polyethylene (BPE) and ordinary concrete were combined to shield the neutron and photon. By using niobium as target backing and water as cooler, it is promising to cope with the problem of thermal damage and hydrogen embrittlement damage. The sizes of moderator and reflector are optimized to have thermal neutron flux as high as possible, while keeping the low ratio of fast neutron flux to thermal neutron flux. The neutron and gamma dose equivalent rates were evaluated and the current shielding configuration is acceptable.

Compilation of Existing Neutron Screen Technology

The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs) by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

The effects of core zoning on optimization of design analysis of molten salt reactor

The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly.

Characterization of Irradiation Sites of Apsara Reactor for $k_{0}$-Based IM-NAA and Its Validation and Application

Three irradiation positions of Apsara reactor were characterized by determining epithermal neutron spectrum shape factor <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex Notation="TeX">$(\alpha)$</tex></formula> and sub-cadmium to epithermal neutron flux ratio <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$(f)$</tex> </formula> for <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">${\rm k}_{0}$</tex> </formula> -based Internal Monostandard Neutron Activation Analysis (IM-NAA). The <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$\alpha$</tex></formula> values were determined by cadmium ratio method using six neutron flux monitors (In, Au, Co, Mo, Zn and Zr). The <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$f$</tex> </formula> -value was also determined by cadmium ratio method using an Au monitor. The <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$\alpha$</tex></formula> values of two irradiation positions are found to be positive, which indicated softened epithermal neutron spectra whereas <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$\alpha$</tex> </formula> value of the third one is negative indicating hardened neutron spectrum. IAEA reference material, RM SL-1 (Lake Sediment) was analyzed by IM-NAA using experimentally determined <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$f$</tex></formula> and <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$\alpha$</tex></formula> values of all three irradiation positions. A total of 16 elemental concentrations were determined using Sc as the internal monostandard and the values are found to be in good agreement with respect to the certified or information values. As an application of IM-NAA, two ceramic archaeological samples along with one soil sample were analyzed using Sc as the monostandard.

Determination of neutron flux parameters in PUSPATI TRIGA Mark II Research Reactor, Malaysia

In standardization NAA, it is necessary to characterize the neutron spectrum parameters such as epithermal neutron flux shape factor (α), thermal to epithermal neutron flux ratio (f), thermal neutron flux (φ th) and epithermal neutron flux (φ epi) in the irradiation facility to determine the concentration of an element in the sample using absolute and k 0 standardization methods. The α and f were determined using Cd-ratio multi monitor method using experimental data obtained in PUSPATI TRIGA Mark II research reactor at four irradiation positions (10, 20, 30 and 40) of the rotary rack. The calculated values of α and f ranged from 0.006 to 0.0281 and 18.56 to 19.12 respectively. The average values of φ th and φ epi were found as 2.33 × 1012 and 1.23 × 1011 n cm−2 s−1 respectively. Moreover, a comparison of the neutron flux parameters in the present study shows an acceptable level of consistency with those of previous studies.

Effect of radial zoning of 241Am content on homogenization of denatured Pu with broad range of neutron energy based on U irradiation test in the experimental fast reactor Joyo

The design consideration of depleted U and Am (DU-Am) oxide fuel pin was performed for Pu denaturing in the framework of the protected plutonium production based on the irradiation analyses of the DU samples irradiated in the environment of broad range of neutron energy in the experimental fast reactor Joyo. From the results of irradiation analyses of DU, it was confirmed that there is a strong dependence of transmutation behavior of DU on the ratio of neutron flux in the resonance region to the total neutron flux, namely resonance neutron ratio, even in a fast reactor. Also, it was confirmed that there is a strong effect of sample material form and shape on generated Pu nuclide inventory especially near the reflector area (>20% resonance neutron ratio), because of the intensive self-shielding of 238U, though less is expected for 241Am. Sensitivity study of hypothetical DU-Am oxide fuel pellet irradiation on neutron energy and burn-up was performed to evaluate significant gradient of radial 238Pu isotopic composition profile (e.g., from 12% to 18% distribution in 3% Am doping, in 30% resonance neutron ratio and in the neutron fluence of 4.0×1022n/cm2 inside a large pellet with softened neutron spectrum), and vulnerability of the fuel pellet surface in terms of Pu denaturing was revealed. Design consideration of radial zoning of 241Am content was introduced to flatten the radial distribution of isotopic composition of Pu. The results of radial zoning of 241Am (4% and 3% Am in the outer and inner zone of DU-Am oxide fuel pellet) in hypothetical irradiation neutronics analysis showed the radial profile of produced Pu is over 15at.% of 238Pu isotopic composition in any zone and meets the criterion of Kimura et al. based on decay heat of Pu to impede utilization to fission explosive devices.

Designing of the 14 MeV neutron moderator for BNCT

In boron neutron capture therapy (BNCT), the ratio of the fast neutron flux to the neutron flux in the tumor (RFNT) must be less than 3%. If a D-T neutron generator is used in BNCT, the 14 MeV neutron moderator must be optimized to reduce the RFNT. Based on the neutron moderation theory and the simulation results, tungsten, lead and diamond were used to moderate the 14 MeV neutrons. Satisfying RFNT of less than 3%, the maximum neutron flux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer, a 14 cm thick lead layer and a 21 cm thick diamond layer.

Characterization of pneumatic fast transfer system irradiation position of KAMINI reactor for k 0-based NAA

The pneumatic fast transfer system position at KAMINI reactor, Indira Gandhi Centre for Atomic Research, India was characterized by determining the epithermal neutron flux shape factor (α) and the sub-cadmium to epithermal neutron flux ratio (f) for k0-based Neutron Activation Analysis (k0-NAA). For determination of α value, bare, Cd-ratio and Cd-cover methods were employed using dual and multi monitors namely Au, Zr and Zn. For calculation of f, Au and Zr monitors were used in the case of bare method and Au monitor was used for cadmium ratio and cadmium cover methods. The determined α-value of PFTS indicated a hard epithermal neutron spectrum and the f value indicated about 96 % thermal neutron component. For validation of k0-NAA method, reference materials namely NIST SRM 1646a (Estuarine Sediment) and BCS Nb-stabilized Stainless Steel (BCS/SS No.261/1) were analyzed. The percentage errors of the determined concentration values of elements were within ±5 % with respect to the certified values and the Z-score values at 95 % confidence level were within ±2 in most of the cases.

Influence of the Neutron Flux Characteristic Parameters in the Irradiation Channels of Reactor on NAA Results Using k&amp;lt;sub&amp;gt;0&amp;lt;/sub&amp;gt;-Standardization Method

An approximation method using to estimate the influence of the uncertainties of the neutron flux characteristic parameters in the irradiation positions on the NAA results using k0-standardization technique was presented. Those are the epithermal reactor neutron spectrum shape-factor α, the effective resonace energy Ε for a given nuclide and the thermal to epithermal neutron flux ratio f. The method is applied to estimate the effect of the uncertainties in the determination of α Ε, and f on final NAA results for some irradiation channels of the Dalat reactor. It also shows that presented method is suitable in practical use for the estimation of the errors due to the uncertainty of the neutron flux characteristic parameters at the irradiation position.

Development and implementation of Høgdahl convention and Westcott formalism for k0-INAA application at Malaysian Nuclear Agency reactor

The present work shows the development of k0-instrumental neutron activation analysis (k0-INAA) method at the Malaysian Nuclear Agency research reactor. To use the k0-INAA method, two formalisms were regulated according to 1/ν and non-1/ν (n, γ) reaction nuclides. The reactor neutron spectrum parameters, the thermal to epithermal neutron flux ratio (f), the epithermal neutron flux shape factor (α) were measured using the bare bi-isotopic monitor and bare triple monitor methods, respectively, based on the Hogdahl convention. In addition, the modified spectral index $$ r(\alpha )\sqrt {T_{n} /T_{0} } $$ , the Westcott $$ g_{Lu} (T_{n} ) $$ factor and the absolute neutron temperature T n parameters were determined using the Westcott formalism. 176Lu was used as non- 1/ν monitor while 197Au, 96Zr and 94Zr were used as 1/ν monitors. The average values of $$ r(\alpha )\sqrt {T_{n} /T_{0} } ,\,g_{Lu} (T_{n} ) $$ and T n were determined to be 0.1795 ± 0.0044, 1.9729 ± 0.0234 and 50.12 ± 3.21°C, respectively. The accuracy of the method was evaluated by analysing IAEA-Soil 7, IAEA-SL 1, NBS SRM 1633A-1 and IAEA-Soil 375 as reference materials. The results show an acceptable level of consistency.

Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

Determination of Neutron Fluxes and Spectrum Shaping Factors in Irradiation Sites of Ghana’S Miniature Neutron Source Reactor (mnsr) by Activation Method After Compensation of Loss of Excess Reactivty

Accurate neutron flux values in irradiation channels of research reactors are very essential to their usage. The total neutron flux of the Ghana Research Reactor-1(GHARR-1) was measured after a beryllium reflector was added to its shim to compensate for excess reactivity loss. The thermal, epithermal and fast neutron fluxes were determined by the method of foil activation. The experimental samples with and without a cadmium cover of 1-mm thickness were irradiated in the isotropic neutron field of the irradiation sites of Ghana Research Reactor-1 facility. The induced activities in the sample were measured by gamma ray spectrometry with a high purity germanium detector. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were taken into account during the experimental analysis. By defining cadmium cutoff energy of 0.55eV, Al-0.1% Au wires of negligible thickness were irradiated at 3kW to determine the neutron fluxes of two irradiation channels, outer channel 7 and inner channel 2 whose Neutron Shaping Factor (α) were found to be (0.037 ± 0.001) and (–0.961 ± 0.034). The neutron flux ratios at the inner irradiation site 2 were found to be, (25.308 ± 3.201) for thermal to epithermal neutrons flux, (0.179 ± 0.021) for epithermal to fast neutrons flux and (4.528 ± 0.524) for thermal to fast neutrons flux, in the outer irradiation site 7, the neutron flux ratios were found to be, (40.865 ± 3.622) for thermal to epithermal neutrons flux, (0.286 ± 0.025) for epithermal to fast neutrons flux and (11.680 ± 1.030) for thermal to fast neutrons flux.

Experimental and MCNP calculations of neutron flux parameters in irradiation channel at Es-Salam reactor

The Algerian research reactor (Es-Salam) is a 15 MW heavy water reactor type, operating since 1992. It became essential to characterize the neutron field in the most useful irradiation positions, in order to guarantee the accuracy in the application of k0-neutron activation analysis (k0-NAA). Experimental value of the thermal to epithermal neutron flux ratio (f) and of the deviation of the epithermal neutron spectrum from 1/E shape (α) were determined using different methods. This work focuses the verification of Monte Carlo neutron flux calculation in typical irradiation channel. Comparison of the results for parameter f obtained experimentally and by Monte Carlo simulations shows good agreement in the irradiation channel studied. The difference between both results is about 2.08%.

Neutron flux parameters for k0-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

Abstract The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (ƒ), epithermal neutron flux shape factor (α), thermal neutron flux (φth) and epithermal neutron flux (φepi) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k0-neutron activation analysis (k0-NAA).The ƒ and α were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of ƒ and α ranged from 33.49 to 47.33 and −0.07 to −0.14, respectively. The φth and the φepi were measured as 2.03 × 1012 (cm−2 s−1) and 6.05 × 1010 (cm−2 s−1) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

Validation of the Monte Carlo Model of the Greek Research Reactor Core

The aim of this study was to validate a Monte Carlo-based model of the Greek Research Reactor-1 (GRR-1) developed with the MCNP5 code. The GRR-1 core was modeled in detail using the exact geometry without approximations. The inventory of the core was derived using the WIMS-ANL code, taking into account the different 235U burnup of each fuel assembly. The model was validated against experimentally determined control rod reactivity worth and neutron flux measurements performed in various irradiation positions. The ratio of the calculated-to-measured integral reactivity of each of the five control rods was found to be 0.972 ± 0.151, 1.083 ± 0.168, 1.156 ± 0.179, 0.874 ± 0.137, and 1.097 ± 0.170. The calculated-to-measured thermal neutron flux ratios ranged from 0.83 ± 0.04 to 1.22 ± 0.07. Therefore, good agreement between MCNP calculated and experimental values was observed. The GRR-1 core model will be fully implemented in the design of material irradiation experiments along with reactor safety and fuel management studies.

Design of the neutral K0L beamline for the KOTO experiment

The J-PARC E14 KOTO experiment aims at the first observation of the KL0→π0νν¯ decay. This experiment requires a well-collimated neutral beam with the small amount of halo neutrons. Collimation lines and beamline components in the beamline have been optimized to suppress halo neutrons by utilizing the GEANT-3 Monte-Carlo simulation package. We have achieved the design with sharp cross-sectional profiles of the beam and the halo-to-core neutron flux ratio of 10−5 level.

Re-characterization of irradiation facilities for k0-NAA at RPI after conversion to LEU fuel and re-arrangement of core configuration

The 1 MW pool-type Portuguese research reactor (RPI) was converted from using highly enriched uranium (HEU) to using low enriched uranium (LEU) fuel in September 2007. Although the new LEU core has the same number of fuel assemblies as the first HEU core, it has three assemblies less than the last HEU core and also has a different arrangement of the beryllium reflectors. Therefore, the primary irradiation positions that are used with the k 0-based neutron activation analysis ( k 0-NAA) at the RPI were redefined and characterized. The neutron spectrum parameters of the deviation of the epithermal neutron spectrum distribution from the 1/ E shape ( α), the thermal to epithermal neutron flux ratio ( f), the thermal to fast neutron flux ratio ( f F) and the neutron temperature ( T n) were determined by using a combination of Au, Lu, Ni and Zr monitors. The experimental results of the neutron spectrum parameters obtained in this study are compared to the results were obtained prior to the refuel conversion in the applicable irradiation positions. The synthetic multi-element standard SMELS was used to verify the accuracy of k 0-NAA applied to practical samples using the newly measured neutron spectrum parameters.