Joel Loridon

New Experimental Results on the Cumulative Yields From Thermal Fission of

The yields of fission products are one of the main characteristics of the fission process. In the field of nuclear waste package characterization, using Photon Activation Analysis (PAA), these yields are needed in order to optimize a technique enabling the identification of actinides (235U, 238U, 239Pu), based on the detection of delayed gamma-rays. As the lack of data in the field of photofission is strongly penalizing for the tuning of this technique, we designed several measurement campaigns in order to determine the yields of various photofission products. The experiments were based on the detection of delayed gamma-rays and delayed neutrons emitted during the same measurement. The feasibility of this technique was first verified in the context of active neutron interrogation, by comparing experimental results for the thermal fission of 235U and 239Pu with reference values provided by several recent databases (ENDFB 6.8, JEFF 3.1). The method was then applied to active photon interrogation, in order to obtain the yields of nuclides formed by the photofission of 235U and 238U. This paper presents the experimental results obtained with these measurements.

^{235} {\rm U}

Abstract The development of a passive-active neutron assay system for alpha low level waste characterization at the French Atomic Energy Commission is discussed. Less than 50 Bq[α] (about 50 μg Pu) per gram of crude waste must be measured in 118-l “European” drums in order to reach the requirements for incinerating wastes. Detection limits of about 0.12 mg of effective 239Pu in total active neutron counting, and 0.08 mg of effective 239Pu coincident active neutron counting, may currently be detected (empty cavity, measurement time of 15 min, neutron generator emission of 1.6 × 108 s-1 [4π]). The most limiting parameters in terms of performances are the matrix of the drum—its composition (H, Cl...), its density, and its heterogeneity degree—and the localization and self-shielding properties of the contaminant.

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Radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant are measured by neutron interrogation with the Differential Die-away measurement Technique (DDT). The purpose is to assay fissile material quantities present in radioactive waste packages. In the future, old hulls and nozzles containing Ion-Exchange Resin (IER) will be measured. IERs provide moderating properties to the matrix, not encountered during the current measurement. In this context, the Nuclear Measurement Laboratory (NML) of the CEA Cadarache has been asked by AREVA NC to explore the possibility of implementing a matrix effect correction method, based on internal monitor (3He proportional counter) signal correlated to the matrix effect. In order to validate this method, a benchmark was performed with PROMETHEE 6 RD10%), in terms of prompt calibration coefficient (useful signal of fissile materials) and internal monitor signal, considering the complexity of the measurement method and numerical model, and the large range of moderator and absorption ratios. The relationship between the prompt calibration coefficient and the internal monitor signal observed in PROMETHEE 6, both for experience and model, can be fitted with a similar function as the industrial measurement cell, the correlation of which being established by numerical simulation. Regressions from experimental and modelling are almost superimposed.

^{239}{\rm Pu}

The cleaning up and dismantling of nuclear facilities lead to a great volume of technological radioactive wastes which need to be characterized in order to be sent to the adequate final disposal or interim storage. The control and characterization can be performed with non-destructive nuclear measurements such as gamma-ray spectrometry. Passive neutron counting is an alternative when the alpha-gamma emitters cannot be detected due to the presence of a high gamma emission resulting from fission or activation products, or when the waste matrix is too absorbing for the gamma rays of interest (too dense and/or made of high atomic number elements). It can also be a complement to gamma-ray spectrometry when two measurement results must be confronted to improve the confidence in the activity assessment. Passive neutron assays involve the detection of spontaneous fission neutrons emitted by even nuclides (238Pu, 240Pu, 242Pu, 242Cm, 244Cm…) and neutrons resulting from (α, n) reactions with light nuclides (O, F, Be…). The latter is conditioned by the presence of high α-activity radionuclides (234U, 238Pu, 240Pu, 241Am…) and low-Z elements, which depends on the chemical form (metallic, oxide or fluorine) of the plutonium or uranium contaminant. This paper presents the recent application of passive neutron methods to the cleaning up of a nuclear facility located at CEA Cadarache (France), which concerns the Pu mass assessment of 2714 historic, 100 litre radioactive waste drums produced between 1980 and 1997. Another application is the dismantling and decommissioning of an uranium enrichment facility for military purposes, which involves the 235U and total uranium quantifications in about a thousand, large compressors employed in the gaseous diffusion enrichment process.

and From Photofission of

To answer safety authority requirements and to optimise the management of radioactive waste produced in retrieval and decommissioning activities, which contains a large variety of matrix materials, the accuracy of neutron measurement techniques has to be continuously improved. Active neutron measurements such as the Differential Die-Away (DDA) technique involving pulsed neutron generator as the neutron source, are widely applied to determine the fissile content of waste packages. Unfortunately, the main drawback of such techniques is coming from the lack of knowledge of the waste matrix composition. Thus, the matrix effect correction for the DDA measurement is an essential improvement in the field of fissile material content determination. Different solutions have been developed to compensate the effect of the matrix on the neutron measurement interpretation for a long time. In Low-Level radioactive Waste (LLW) packages examination, the most widely used methods are based on neutron flux monitoring using small 3He proportional counters added inside the detection device and associated to the “Matrix Interrogation Source” (MIS) measurement. This technique was originally developed for passive neutron measurement. It needs a specific measurement step which can be operated with the neutron generator or, most of the time, with an external isotopic neutron source such as 252Cf located as closed as possible to the waste drum. This step represents a limiting factor for the examination management and duration. In this context, this paper describes a new approach developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time. This is a major objective in the Non Destructive Assay (NDA) especially to enhance industrial process efficiency of large number of waste packages inspection. It deals with an innovative matrix correction method for radioactive waste embedded in a large variety of matrices regarding the density range (0.07 – 0.9 g.cm−3) as well as the composition (wood representative of hydrogenized matrix, PVC, iron, etc.). The implementation of this method is based on the analysis of the raw signal with an optimisation algorithm called the simulated annealing algorithm. This algorithm needs a reference data base of Multi-Channel Scaling (MCS) spectra, to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals as representative as possible of the real measurement conditions. This database has been provided by a set of active signals from experimental matrices (mock-up waste drums of 118 litres) recorded in a specific device dedicated to neutron measurement research and development of the Nuclear Measurement Laboratory of CEA-Cadarache, called PROMETHEE 6. This equipment has been designed to reach an empty cavity detection efficiency of 25%. It is equipped with a pulsed (D-T) neutron generator which can reach an average neutron emission rate up to 2.4 109 ns−1 with a pulse duration of 200 µs. This high technology performance allows achieving very low detection limits with the classical DDA measurement of fissile matter located in light waste matrices (close to 30 µg of 239Pu with an active total measurement time of 900 s). The simulated annealing algorithm is applied to make use of the effect of the matrices on the total active signal of DDA measurement. Furthermore, as this algorithm is directly applied to the raw active signal, it is very useful when active background contributions can not be easily estimated and removed. Most of the cases tested during this work which represents the feasibility phase of the method, are within a 4% agreement interval with the expected experimental value. Moreover, one can notice that without any compensation of the matrix effect, the classical DDA prompt neutron signal analysis may induce an underestimation of more than a factor of 200 on the fissile mass determination for the cases tested in this study. The unexpected so good agreement is a very promising result for the method knowing that the compositions of the mock-up drums are quite representative of the most frequently encountered matrices in LLW packages. This work is the first step of a more general thought carried out to increase the relevance of the whole treatment of DDA measurements from innovative electronic tools (specific fast charge amplifiers, list mode data card system…) up to optimised home made algorithms developed for the post-treatment of the measurements recorded by the list mode data card system.