Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

International audienceMetallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor.

International audienceCABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155bar, 300°C). This project lied within a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimental program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration.During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment

Localization of nuclear materials in large concrete radioactive waste packages using photofission delayed gamma rays

International audienceThe characterization of radioactive waste packages is mandatory for their transport, interim storage and final disposal. In this framework, the Nuclear Measurement Laboratory of CEA DES IRESNE Institute, at Cadarache, France, uses a high-energy electron linear accelerator (LINAC) to produce an interrogating bremsstrahlung beam with endpoint energies ranging from 9 to 21 MeV to perform X-ray imaging and high-energy photon interrogation on large concrete packages. In particular, highenergy photon beam induces photofission reactions in both fissile (235U, 239Pu, 241Pu) and fertile (238U, 240Pu, 232Th, etc.) actinides possibly present in the radioactive waste. In order to assess their mass, we use delayed gamma rays emitted by their photofission products, which are measured with a 50 % relative efficiency High-Purity Germanium (HPGe) detector. Actinide differentiation, which is important for the fissile mass estimation, is based on the ratios of gamma rays emitted by different photofission products and requires appropriate corrections for the gamma attenuation in concrete. To this aim, we report here a localization method of point-like nuclear materials in the concrete matrix, based on the differential attenuation of several gamma rays emitted by a same photofission product. We use here the 1435.9 and 2639.6 keV lines of 138Cs, with both experimental data and MCNP numerical simulations to determine the (r,θ) coordinates of nuclear materials. Then, the depth inside the concrete matrix, which is determined with a precision of a few percent, mainly depending on counting statistics on 1435.9 and 2639.6 keV net peak areas, is used to correct for the different gamma ratios used in the actinide identification method. Experimental tests with uranium samples have been performed to validate the localization method.Key words: Photofission / Uranium / Delayed gamma rays / Bremsstrahlung / MCNP / Concrete matrix / Nuclear material localizatio

Status of the nuclear measurement stations for theprocess control of spent fuel reprocessing at AREVA NC/La Hague

International audienceNuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectrometry, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality-safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remaining after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction in stainless steel containers. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations n°1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station n°1 allows determining the burn-up of the irradiated fuel by gamma-ray spectrometry with HP Ge (high purity germanium) detectors. The burn-up is correlated to the $^{137}$Cs and $^{134}$Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station n°3 is dedicated to the control of the correct fuel dissolution, which is performed with a $^{137}$Cs gamma ray measurement with a HP Ge detector. Station n°7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional $^3$He detectors. So far, large campaigns of reprocessing of the UOX fuels with a burn-up rate up to 60 GWd/t have been performed at AREVA/La Hague. This paper presents a brief overview of the current status of the nuclear measurement station

Nuclear data production, calculation and measurement: a global overview of the gamma heating issue

The gamma heating evaluation in different materials found in current and future generations of nuclear reactor (EPRTM, GENIV, MTR-JHR), is becoming an important issue especially for the design of many devices (control rod, heavy reflector, in-core & out-core experiments…). This paper deals with the works started since 2009 in the Reactor Studies Department of CEA Cadarache in ordre to answer to several problematic which have been identified as well for nuclear data production and calculation as for experimental measurement methods. The selected subjects are: Development of a Monte Carlo code (FIFRELIN) to simulate the prompt fission gamma emission which represents the major part of the gamma heating production inside the core Production and qualification of new evaluations of nuclear data especially for radiative capture and inelastic neutron scattering which are the main sources of gamma heating out-core Development and qualification of a recommended method for the total gamma heating calculation using the Monte Carlo simulation code TRIPOLI-4 Development, test and qualification of new devices dedicated to the in-core gamma heating measurement as well in MTR-JHR as in zero power facilities (EOLE-MINERVE) of CEA, Cadarache to increase the experimental measurement accuracy