[en] The Isotope Dilution Mass Spectrometry (IDMS) measurement technique provides a means for determining the unknown amount of various isotopes of an element in a sample solution of known mass. The sample solution is mixed with an auxiliary solution, or tracer, containing a known amount of the same element having the same isotopes but of different relative abundances or isotopic composition and the induced change in the isotopic composition measured by isotope mass spectrometry. The technique involves the measurement of the abundance ratio of each isotope to a (same) reference isotope in the sample solution, in the tracer solution and in the blend of the sample and tracer solution. These isotope ratio measurements, the known element amount in the tracer and the known mass of sample solution are used to calculate the unknown amount of one isotope in the sample solution. Subsequently the unknown amount of element is determined. The purpose of this paper is to examine the optimization of the ratio of the estimated unknown amount of element in the sample solution to the known amount of element in the tracer solution in order to minimize the relative uncertainty in the determination of the unknown amount of element

[en] A high-speed gamma-ray spectroscopy system for verification of nuclear fuel of the BWR type has been constructed. It uses a gamma-ray collimator, commercial high-speed NIM electronics, a fast PC and especially designed software and PC interface. It allows accurate determination of e.g. ¹⁵⁴Eu in presence of strong activities of ¹³⁴Cs and ¹³⁷Cs. With this technique isotopic correlations may be used for independent determination of fuel parameters

[en] Enriched uranium fuel assemblies were measured with an Active Well Coincidence Counter (AWCC) at the Beloyarskaya Nuclear Power Plant. Special AWCC inserts, electronics, and software were used. Stationary and scanning measurements were performed to establish calibrations and performance specifications for the assay of ²³⁵U and ²³⁵U/cm for BN600 fuel

[en] The improvement of accuracy of nuclear physics data of transactinium nuclides, namely half-lives determination, is of vital importance for the development of nuclear physics assay and monitoring methods and ensuring their reliability. The half-life of ²⁴⁰Pu due to α radiation has been determined to be 6563+7 years. The value of ²⁴⁰Pu half-life due to spontaneous fission has not been precisely determined until recently. It varied from 1.15x10¹¹ to 1.47x10¹¹ years. The value of ²⁴⁰Pu spontaneous fission half-life was obtained using results of precision neutron flux measurements from two spherical metal plutonium samples with ²⁴⁰Pu content of about 90% and impurities less than 0.3%. The effect of self-multiplication specific neutron fluxes of (α,n) reactions and spontaneous fission of ²³⁸ Pu and ²⁴²Pu were taken into account. The obtained value of ²⁴⁰Pu spontaneous fission half-life consists of (1.15+0.02)x10¹¹ years

[en] The application of new video surveillance electronics to safeguards has introduced an urgent need to formulate and adopt video standards that will ensure the highest possible video quality and the orderly introduction of data insertion. Standards will provide guidance in the application of image processing and digital techniques. Realistic and practical standards are a benefit to the IAEA, Member States, Support Programme equipment developers and facility operators, as they assist in the efficient utilisation of available resources. Moreover, standards shall provide a clear path for orderly introduction of newer technologies, whilst ensuring authentication and verification of the original image through the video process. Standards emerging from IAEA are an outcome of experience based on current knowledge, both within the safeguards arena and the video parent industry which comprises commercial and professional television. This paper provides a brief synopsis of recent developments which have highlighted the need for a surveillance based video standard together with a brief outline of these standards

[en] As one of the most important industrial electronuclear states in the world, with a comprehensive national fuel cycle (from mines to reprocessing plants) on the one hand, and as a nuclear supplier as well as a nuclear weapon state on the other hand, France deemed it necessary to establish its own domestic safeguards system, in order to struggle against hazards of malevolence at the national level and to contribute efficiently to nuclear non-proliferation at the international level. The French legal regulations are based on the main following principles: (1) Licensing of activities related to import-export, to the possession or to the transportation of nuclear material; (2) Control performed by the operator, including NMCA, C/S an physical protection; (3) Inspection by the domestic Safeguards Authority; (4) Penalties. Specificities of domestic and international Safeguards (Euratom, IAEA) concerning objectives and scopes, information required, accounting, inspection types, R and D are illustrated and possible interactions are considered

[en] A number of reasons are given why more isotope-specific or isotope-related thinking should occur in measurements for Safeguards and Nuclear Material Management. It is isotopes which are fissile. Isotope-specific methods are available for direct quantitative determination of isotopes without a detour via elemental content with its problems in sample stoichiometry. Fissile isotopes in input solutions of reprocessing plants are already assayed by an isotope-specific method (IDMS*) and enable direct isotope-accountancy. Within the same amount of element, fissile isotopes (e.g. ²³⁵U) can be substituted for non-fissile isotopes (e.g. ²³⁸U) using isotope separation techniques of the near future leaving the total amount of element the same but changing the Safeguards character of the material drastically. Isotope-specific ratio of ratio measurement approaches, applied to IDMS have the potential of breaking through the 0.1% uncertainty barrier, at the same time making measurements easier. Large dynamic ratio measurement capability would enable the determination of the total amount of fissile isotope within a nuclear material batch (e.g. an input tank) directly against a small (ppm) but known added amount of a spike isotope (e.g. ²³³U) without the necessity of separate calibration of the tank for mass or volume of solution

[en] The interest in the safeguards of fissile material focuses on a limited number of compounds which play key roles in the nuclear fuel cycle. Amongst these materials Uranium Dioxide pellets are of considerable importance as they enter the reactors in order to generate energy. In LWR's pellets with an initial ²³⁵U content of about 3 mass % are used, whereas natural or depleted material is applied for the breeding zone in FBR's. The 89/90 round o REIMEP covered Uranium materials with ²³⁵U abundances in the range of natural or depleted material. UO₂ pellets were distributed to 21 laboratories for analysis. The participating laboratories were asked to determine the Uranium content and the isotopic composition of the material. The results reported by the participants are presented as graphs thus giving a picture of the state-of-the-practice

[en] An highly enriched ²⁴⁴Pu isotopic reference material (CBNM IRM-042a) has been prepared and certified for ²⁴⁴Pu isotope concentration. The certified value of (2.257 7 ± 0.004 4).10¹⁸ atoms ²⁴²Pu.kg^-1 of solution has been established by isotope dilution mass spectrometry. The plutonium isotopic composition has been determined by thermal ionization mass spectrometry and calibration of these measurements by means of synthetic ²⁴²Pu/²³⁹Pu mixtures. The isotopic reference material is supplied in a sealed glass ampoule containing approximately 10 g of a 5M nitric acid solution at an approximate concentration of 1 μg Pu per g solution. This isotopic reference material is part of a systematic CBNM programme to supply spike isotopic reference materials of various isotopes at different concentrations

[en] The programme AECC (Active Euratom Coincidence Counters) has been developed at the Joint Research Center, Ispra by the Euratom Safeguards Directorate, Luxembourg and the Safety Technology Institute, Ispra for the acquisition, evaluation, management and storage of measurement data originating from active neutron interrogation of HEU and LEU fuel materials. The software accommodates the implementation of the NDA (Non Destructive Assay) procedures for the Active Well Coincidence Counters and Active Neutron Coincidence Counters deployed by the Euratom Safeguards Directorate, Luxembourg