Performance test of the Multi-Channel Analyzer MCA-527 for Nuclear Safeguards Applications Test in the PERLA Laboratory at JRC, Ispra

The multi channel analyzer MCA-166 of GBS Rossendorf GmbH, Germany, has been a base instrument for gamma spectrometry both of IAEA and EURATOM for nuclear safeguards applications. Since essential internal electronic chips of the MCA-166 are not provided any more IAEA and the German support program to the IAEA decided to endorse the development of a follower instrument, the MCA-527, with the same company. The performance of this new instrument was tested with respect to parameters, which are essential for safeguards applications: Dead time correction for U enrichment measurements, peak shape for high resolution applications MGA and MGAU, high count rate performance with CdZnTe detectors for spent nuclear fuel, temperature stability of the MCA itself. The tests cover all important detector types applied by IAEA and EURATOM: NaI, NaI with internal Am source, planar Ge, coaxial Ge, LaBr3, and CdZnTe detectors. The tests were made with nuclear materials U and Pu, and with 137Cs and 60Co to simulate spent nuclear fuel. They cover count rate ranges up to about 70 000 … 100 000 cps for U and Pu and with CdZnTe detectors up to 300 000 cps. The report provides a series of setup files for different detector types. The result of the test is: The performance of the MCA-527 meets the functional requirements for gamma spectrometric measurements for nuclear safeguards applications. Its parameters are as good as or the ones of the MCA-166 or superior to them. The MCA-527 can also be used for neutron measurements in List mode. Its performance for neutron counting will be described in separate report.JRC.E.8-Nuclear securit

Survey of State-of-the-art NDA Methods Applicable to UF6 Cylinders - IAEA Task n 07/TAU-04

In the framework of a project aiming to establish an unattended measurement station at an isotope enrichment facility, IAEA required a study to describe the state of the art of NDA methods applicable to UF6 cylinders. The objective of the present work is to provide a feasibility assessment study of all known NDA techniques applicable to the quantitative verification of all uranium categories involved in an enrichment processing plant. The quantification of the UF6 cylinders covers: - the determination of the enrichment, - the confirmation the UF6 mass ( assumed to have been previously weighted by the plant operator and independently verified by inspectors), - the assay of the UF6 homogeneity. The different hypothesis and practical constraints to be taken into account for the study requirements are [1]: - the cylinders to be considered are either 30B type ( product) or 48Y type ( feed and tail), - the enriched uranium is either from natural origin or reprocessed uranium, - the cylinders must be assayed at various temperatures, - the distance between the cylinder and the detector must be at least 50 cm to allow for safe movements of the cylinders, - the UF6 mass determination would be accurate within 10% for low enriched uranium, 15% for natural uranium and 20% for depleted uranium, - the enrichment determination must be given with a total uncertainty which does not excess: ¿ 4.5% for low enriched uranium product, ¿ 9.5% for natural uranium, ¿ 18% for depleted uranium, - the measurements have to be performed in 5 minutes and in remote mode to minimize the intrusion on normal plant operator. With the objectives and assumptions as described above in mind, this document first gives an overview of the radiation properties of UF6 (chapter A) as well as some practical considerations regarding the 48Y and 30B cylinders (chapter B). The next part reviews the classical NDA methods applicable to UF6 and refers to intense measurement campaigns carried out in the years 70 -80 (chapter C), whereas the chapter D is dedicated to specific studies involving more recent techniques such as analysis of delayed neutrons and delayed photons. The most appropriate techniques will be then investigated in chapter E. The study will be based on our own results of previous measurement campaigns (235U determination with gamma detectors with germanium or LaBr3 detectors) and on MCNP simulations (passive and active neutron methods).JRC.G.8-Nuclear securit

Determination of 242Pu Abundance in Plutonium Samples Using Self-fluorescent X-ray Emission

Non-destructive gamma spectrometric measurements of the isotopic composition of plutonium are routinely carried out in the framework of nuclear safeguards inspections. High-resolution spectra of Pu obtained from planar Ge detectors are de-convoluted by means of special codes such as MGA or FRAM and these results are then applied when determining the total plutonium mass using neutron coincidence measurement techniques. The radiation emission of 242Pu is so weak that its abundance cannot be measured directly; hence isotopic correlations are utilized to estimate it in low and medium burn-up Pu material. The present work investigates the feasibility of determining the total Pu/Am ratio, exploiting X-ray fluorescence lines present in the plutonium spectrum. Combining the total Pu/Am ratio with individual Pu isotopic ratios to 241Am then allows an estimate of the 242Pu to be made. To this end, reference samples of plutonium, with an 241Am content comprised between 5 and 7 %, were measured using a planar germanium detector. The spectra were processed using MGA followed by quantification of the X-ray fluorescence peak of Americium, and then further calculations, including MCNP simulations. The conclusion of the present work is that, when the americium content of the sample is high enough to produce an Am K1 line which can be exploited, it is possible to indirectly determine the 242Pu abundance by analysing the X-ray fluorescent lines present in the plutonium gamma spectrum, thus raising the prospect to apply this method to cases for which the algorithms based on empirical correlations cannot be applied.JRC.G.II.7-Nuclear securit

Attenuation of a Non-Parallel Beam of Gamma Radiation by Thick Shielding. Application to the Determination of the 235U Enrichment with NaI Detectors.

Abstract not availableJRC.G-Institute for the Protection and the Security of the Citizen (Ispra

Monte Carlo Modelling of a n-type coaxial high purity germanium detector

A modelling of a N-type coaxial germanium detector was performed with the Monte Carlo code MCNP to calculate the net peak areas in the spectra in the energy range from 50 to 1500 keV. Since the computed values based on manufacturer data on the detector deviate significantly (up to 60 %) from the experimental data, careful scanning measurement across the front area and the side of the detector were carried out using collimated photon beams in order to investigate the reason of the efficiency deficit. As a result, a much smaller active detector volume than described by the manufacturer was measured and the detector response over the crystal length was inhomogeneous. After adjustment of the detector dimensions in the MCNP modelling, the deviations between calculated and empirical were found to be less than 4% for the whole energy range.JRC.E.8-Nuclear securit

Determination of the Uranium Enrichment with the NaIGEM Code.

Abstract not availableJRC.G-Institute for the Protection and the Security of the Citizen (Ispra

Handbook of gamma spectrometry For Non-destructive Assay of Nuclear Materials

This document forms the fourth edition of the report untitled "Handbook of Gamma Spectrometry methods for Non-Destructive Assay of nuclear materials" (EUR 19822 EN). It has been updated from the June 2006 release. The organization by chapters has remained unchanged, but the content has been revised. Were added to the contents: -spectrum pictures, -new data, -new measurement results making references to results from our laboratory, -new links providing an easy navigation through the glossary and annexes, -new references. Clearly, some minor mistakes of the third version were corrected, but avoided adding too much new information, since the booklet should stay restricted to inspector needs.JRC.DG.G.8-Nuclear securit

Measurement of 235U Enrichment with a LaBr3 Scintillation Detector

This paper describes the performance of a 1.5 * 1.5 inch LaBr3 gamma radiation detector for determining the 235U enrichment by non destructive analysis. The spectrometric properties of the detector, brought to market under the trade name BrillanCe-380 [1] were first evaluated. Enrichment measurements were subsequently carried out on certified uranium samples with enrichment ranging from 0.31% to 60% and on UF6 containers of the type 30B and 48Y in different experimental conditions.JRC.DG.G.8-Nuclear securit

Experience with Nuclear Inspector Training at JRC, Ispra

About 500 nuclear safeguards inspectors are working at the IAEA, EURATOM and as national inspectors in Europe. Up to 50 of them are recruited every year and need training for their new work, comprising all its aspects. A higher number of inspectors need refreshment courses or introductions into new working fields. Moreover, new instruments or techniques require special training, in class, laboratory or in field. This presentation is based mainly on experience with laboratory courses on nuclear measurements for the verification of nuclear material.JRC.E.8-Nuclear securit