Non-destructive assay and computational model for enrichment verification of UF6 cylinders

El‐Mongy, Sayed A.; Allam, Kh A; Farid, Osama M.

doi:10.1016/j.radmeas.2007.12.014

Cited by 9 publications

(8 citation statements)

References 2 publications

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“…Gamma-ray spectroscopy has the advantages of portability [57] and non-intrusiveness for on-site measurements; however, the signal response is known to be heavily dependent on cylinderrelated parameters such as variation in wall thickness and filling profile [60,61]. For example, it has been stated that typical variation of the wall thickness (~ 0.5 mm) from the nominal value easily leads to an enrichment error of 6% [60].…”

Section: Gamma-ray Spectroscopymentioning

confidence: 99%

“…Furthermore, the mean free path for the dominant -ray at 185.7 keV from 235 U is only ~ 2 mm [62] due to self-absorption, which makes this technique unfavorable for assaying the inner volume of the UF 6 cylinder and unable to detect certain diversion scenarios [63]. A recent report [61] studied -ray spectroscopy on UF 6 cylinders, with correction for attenuation of the 235 U -ray due to the cylinder wall. For a cylinder certified with 19.75% of 235 U, the reported enrichment was 18% (i.e., 9% relative bias) [61].…”

Section: Gamma-ray Spectroscopymentioning

confidence: 99%

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Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

Chan¹,

Valentine²,

Russo³

2017

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The primary objective of the present study is to identity the most promising, viable technologies that are likely to culminate in an expedited development of the next-generation, field-deployable instrument for providing rapid, accurate, and precise enrichment assay of uranium hexafluoride (UF 6 ). UF 6 is typically involved, and is arguably the most important uranium compound, in uranium enrichment processes. As the first line of defense against proliferation, accurate analytical techniques to determine the uranium isotopic distribution in UF 6 are critical for materials verification, accounting, and safeguards at enrichment plants. As nuclear fuel cycle technology becomes more prevalent around the world, international nuclear safeguards and interest in UF 6 enrichment assay has been growing.At present, laboratory-based mass spectrometry (MS), which offers the highest attainable analytical accuracy and precision, is the technique of choice for the analysis of stable and longlived isotopes. Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched UF 6 at declared facilities by collecting a small amount (between 1 to 10 g) of gaseous UF 6 into a sample bottle, which is then shipped under chain of custody to a central laboratory (IAEA's Nuclear Materials Analysis Laboratory) for high-precision isotopic assay by MS. The logistics are cumbersome and new shipping regulations are making it more difficult to transport UF 6 . Furthermore, the analysis is costly, and results are not available for some time after sample collection. Hence, the IAEA is challenged to develop effective safeguards approaches at enrichment plants. In-field isotopic analysis of UF 6 has the potential to substantially reduce the time, logistics and expense of sample handling. However, current laboratory-based MS techniques require too much infrastructure and operator expertise for field deployment and operation. As outlined in the IAEA Department of Safeguards Long-Term R&D Plan, 2012-2023, one of the IAEA long-term R&D needs is to "develop tools and techniques to enable timely, potentially real-time, detection of HEU (Highly Enriched Uranium) production in LEU (Lowly Enriched Uranium) enrichment facilities" (Milestone 5.2). iv measurements that do not require the stated high accuracy or precision of the ITVs to which it was compared).The list presented below summarizes the outcome of the present study, and groups all reviewed techniques into "recommended", "promising", or "not recommended"; the "benchmark" techniques are also included and labeled as such. The list is presented in a highly abridged way such that only the final recommendations are given. The performance of a technique in the seven evaluation metrics and its ranking are summarized in Table 6.1, whereas short discussion and comments on the recommendation can be found in Section 6.2 of this report. Furthermore, comprehensive and in-depth reviews on the scientific principle of each technique, its instrument option, its limitations and main gaps betwe...

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Section: Gamma-ray Spectroscopymentioning

confidence: 99%

Section: Gamma-ray Spectroscopymentioning

confidence: 99%

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

Chan¹,

Valentine²,

Russo³

2017

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“…The International Atomic Energy Agency safeguards inspection, and verification relies on accurate measurements of the fissile content at nuclear facilities. [3][4][5][6][7][8] Neutron measurements are typically used to determine the mass of plutonium and uranium in fresh fuel. 2 NDA methods (neutron and gamma measurements) are used to detect and estimate the mass of fissile content in the broad range of material matrix throughout the nuclear fuel cycle.…”

Section: Introductionmentioning

confidence: 99%

“…2 NDA methods (neutron and gamma measurements) are used to detect and estimate the mass of fissile content in the broad range of material matrix throughout the nuclear fuel cycle. [3][4][5][6][7][8] Neutron measurements are typically used to determine the mass of plutonium and uranium in fresh fuel. Neutron measurements are used either for total counting (ie, pure metal fissile samples) or coincidences/multiplicity counting (ie, impure samples whose [α, n] to spontaneous fission (SF) ratio is well known or impure fissile samples in general).…”

Section: Introductionmentioning

confidence: 99%

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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“…The detection of fissile materials and its mass assay are of primary interest for nuclear security and nuclear safeguards. [3][4][5][6][7] The gamma rays associated with the decay from both 235 U and 238 U, which provide highly accurate information about nuclear materials are used for assay. [6,8,9] This study is mainly highlight on the assay of uranium content assay of leashed and purified yellow cake samples of be 40.5 and 67.5 % before and after purification respectively.…”

Section: Introductionmentioning

confidence: 99%

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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Non-destructive assay and computational model for enrichment verification of UF6 cylinders

Cited by 9 publications

References 2 publications

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

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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