Passive Non-Destructive Assay based on gamma-ray spectrometry to verify UO2 samples in the form of powder and pellet

Tohamy, M.; El-Ghany, Sahar Abd; El-Minyawi, S.M.; Fayez-Hassan, M.; El-Hakim, Elham H.; El‐Mongy, Sayed A.; Comsan, M.N.H.

doi:10.1016/j.anucene.2015.09.001

Cited by 21 publications

(16 citation statements)

References 9 publications

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“…The experimental arrangement of the facility including the neutron irradiator, neutron house as well as the neutron source handling system is illustrated in Figure. 1. Measurement procedures are described in a previous paper [1].…”

Section: Experimental Facility Descriptionmentioning

confidence: 99%

“…In this work, the neutron irradiation facility designed and constructed previously in the shape of a spherical container was used [1]. The design aimed to achieve symmetry with respect to the position of the neutron source used at the center of the sphere.…”

Section: Experimental Facility Descriptionmentioning

confidence: 99%

“…The thermal and epithermal fluxes were measured for different combinational setups to determine the best position for maximum thermal neutron flux for each setup [1,11]. The epithermal and thermal fluxes are calculated according to equations (7) and (9) for both In and Au samples, respectively.…”

Section: Measured Neutron Fluxesmentioning

confidence: 99%

“…A neutron house was used to accommodate the neutron source outside the facility [1]. The shield of the neutron house was designed such that the dose rate at the outside surface of the shield did not exceed the dose limits given according to ICRP-60 [5].…”

Section: Experimental Facility Descriptionmentioning

confidence: 99%

“…In the NAA the sample to be analyzed is exposed to a thermal neutron flux; then, the induced activity of the sample is measured using absolute or comparative method. For this purpose a previously described small-sized neutron irradiator prototype based on a 5 Ci (1.85×10 11 Bq) Am-Be isotopic neutron source was prepared [1]. The use of the neutron irradiator presents the advantage of supplying a stable neutron flux over a long period [2][3][4], thus eliminating the need for using expensive neutron producing devices such as reactors or accelerators.…”

Section: Introductionmentioning

confidence: 99%

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Characterization of Neutron Field in a Spherical Irradiation Facility

Tohamy¹

2015

AJMP

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Abstract:A small-sized dual-hemisphere irradiation facility designed and constructed for using Am-Be isotopic neutron source to study and characterize the neutron field around sample irradiation position for obtaining maximum thermal neutron flux. The foil activation method based on 115 In and 197 Au was used for monitoring the thermal neutron flux. For field characterization, several combinations of moderation media have been tested (air, sand and water) aiming at the identification of the best combinational setups for sample irradiations using thermal neutrons. To account for epithermal contribution, reaction rates and flux calculations were corrected using tabulated values of Westcott g-factors and resonance integrals for these media. The optimal position for sample irradiation using thermal neutron for both water-water and water-sand setups, inside the irradiation channel, was found at about 6.9 to 10.9 cm from the center of the neutron source.

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Section: Experimental Facility Descriptionmentioning

confidence: 99%

Section: Experimental Facility Descriptionmentioning

confidence: 99%

Section: Measured Neutron Fluxesmentioning

confidence: 99%

Section: Experimental Facility Descriptionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Characterization of Neutron Field in a Spherical Irradiation Facility

Tohamy¹

2015

AJMP

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Computation of the neutron multiplicity moments for research reactor fuels using MCNP6 and SOURCES4c

Dim

Aghara

2019

Int J Energy Res

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Summary Fissile material detection and quantification are often necessary for safeguards, nuclear security, and fuel management. Nondestructive assay, neutron, and gamma measurements are reliable means, which can facilitate the detection and estimation of the mass of fissile materials in a broad range of material matrix. Various flavours of neutron measurement are routinely used by facilities (like nuclear reactors, enrichment, and fabrication plants) to quantify fissile material mass and inventory lists. The Monte Carlo code, MCNP6, is used to model several neutron multiplicity measurements. A simulation scenario is set up in MCNP6 using the JCC71 neutron slab counter to obtain the multiplicity moments for fresh and irradiated fuel assemblies from the UMass Lowell Research Reactor (UMLRR) and Worcester Polytechnic Research Reactor (WPIRR). An MCNP6 burnup is initially performed on the fuel types under study to generate used fuel isotopic. The fresh and or used fuel isotopic is then used to produce independent SOURCES4c input tape1 files. SOURCES4c is used to generate (α, n), spontaneous fission spectrum, and the associated neutron emission rates necessary for the various fixed fuel source definitions in MCNP6 calculations. Under the comprehensive safeguards agreements, the International Atomic Energy Agency has the right and obligation to verify that no nuclear material is diverted from peaceful use to nuclear weapons or other nuclear explosive devices. Research reactors are required to be safeguarded facilities under the comprehensive safeguards. Several research efforts have studied the various flavours of neutron measurement for commercial power reactor operating at high power and long burnups; however, not nearly as many studies have been performed with neutron measurements for research reactors operating at relatively lower power and have significantly lower burnup. This work looks to establish the relevant isotopes to overall neutron source rate as well as the process involved in performing a typical neutron multiplicity measurement simulation for a research reactor fuel. The results demonstrate that the single and double moments for Worcester Polytechnic Institute (WPI) and UMLRR fuels can be measured reliably using two JCC71 slab detectors. The moment for the UMLRR and WPIRR fuel (in both fresh and used states) was estimated with a relative error below 0.031 for singles and 0.081 for doubles. The two fresh fuel types cannot be differentiated from each other on the sole basis of neutron analysis. However, fresh and irradiated fuel can be distinguished based on neutron multiplicity measurements.

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Ore Leashing Processing for Yellow Cake Production and Assay of their Uranium Content by Radiometric Analysis

Abdel‐Rahman

El‐Mongy

2017

Zeitschrift anorg allge chemie

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Abstract. In this study, Ore granite samples were collected from Gattar site for leashing of yellow cake. The process involves heap leaching of uranium through four main steps; size reduction, leaching, uranium purification, and finally precipitation and filtration. The separation process has been given in details and as flow chart. Gamma spectrometry based on HpGe detector and energy dispersive X-ray (EDX) were used to assay uranium content and activity before and after separation. The uranium weight percentage value as measured by EDX were found to

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Passive Non-Destructive Assay based on gamma-ray spectrometry to verify UO2 samples in the form of powder and pellet

Cited by 21 publications

References 9 publications

Characterization of Neutron Field in a Spherical Irradiation Facility

Characterization of Neutron Field in a Spherical Irradiation Facility

Computation of the neutron multiplicity moments for research reactor fuels using MCNP6 and SOURCES4c

Ore Leashing Processing for Yellow Cake Production and Assay of their Uranium Content by Radiometric Analysis

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