Christian Passard

The help of simulation codes in designing waste assay systems using neutron measurement methods: Application to the alpha low level waste assay system PROMETHEE 6

Abstract The design of a specific nuclear assay system for a dedicated application begins with a phase of development, which relies on information from the literature or on knowledge resulting from experience, and on specific experimental verifications. The latter ones may require experimental devices which can be restricting in terms of deadline, cost and safety. One way generally chosen to bypass these difficulties is to use simulation codes to study particular aspects. This paper deals with the potentialities offered by the simulation in the case of a passive–active neutron (PAN) assay system for alpha low level waste characterization; this system has been carried out at the Nuclear Measurements Development Laboratory of the French Atomic Energy Commission. Due to the high number of parameters to be taken into account for its development, this is a particularly sophisticated example. Since the PAN assay system, called PROMETHEE (prompt epithermal and thermal interrogation experiment), must have a detection efficiency of more than 20% and preserve a high level of modularity for various applications, an improved version has been studied using the MCNP4 (Monte Carlo N-Particle) transport code. Parameters such as the dimensions of the assay system, of the cavity and of the detection blocks, and the thicknesses of the nuclear materials of neutronic interest have been optimised. Therefore, the number of necessary experiments was reduced.

Alpha-particle low-level waste control : Improvement of the promethee 6 assay system performances

Abstract The PROMpt, Epithermal and THErmal interrogation Experiment, version 6 (1996) (PROMETHEE 6) assay system for alpha-particle low-level waste characterization, developed for research and development purposes, includes both passive and active neutron measurement methods. Developed at the Commissariat à l’Energie Atomique, Cadarache Centre, in cooperation with COGEMA, its aim is to reach the incinerating alpha-particle waste requirements (<50 Bq[α]/g of crude waste, i.e., ~50 μg of Pu per drum) in 118-l “European” drums (460 mm in diameter and 750 mm high). Good preliminary results were presented: detection limits of ~0.12 mg of effective 239Pu in total active neutron counting and 0.08 mg of effective 239Pu in coincident active neutron counting [empty cavity, measurement time of 15 min, neutron generator emission of 1.6 × 108 s-1 (4π)]. Those results are improved with the use of a higher neutron source emission [GENIE 36 generator, neutron emission of 2.4 × 109 s-1 (4π)] and working on the configuration of the detector units. In the total counting mode, the gain is a factor of ~4 in a cellulose matrix and 3.1 in a polyvinyl chloride matrix. In the coincidence counting mode, these factors are 1.8 and 1.7, respectively. After a very short description of PROMETHEE 6, this paper presents the last and best performances that were obtained with the increased neutron source. Studies on the detection limit variations with the use of borated shields in front of the detection units and around the neutron generator also are dealt with.

Experimental Qualification With a Scale One Mock-Up of the “Measurement and Sorting Unit” for Bituminized Waste Drums

Within the framework of the cleaning operation of the Marcoule reprocessing plant UP1 (France), the CEA (French Atomic Energy Commission) developed a measurement system for 225-liter drums filled with bituminized radioactive sludge originating from the effluent treatment. This work was carried out for the CODEM, which is an economic interest group made up of CEA, EDF (the French public utility) and COGEMA (the operator of UP1). CODEM is in charge of UP1 dismantling operations, especially waste retrieval. The bituminized waste drums mainly contain plutonium, americium, uranium, curium and various beta emitters among which some are responsible for significant gamma irradiation, such as 137 Cs. The aim of this system is to sort the packages according to their radioactive level, so as to direct them towards the French Aube Center, which is a surface repository. This means they must meet the acceptance criteria related to their activities. Otherwise, they will remain in interim storage in Marcoule, pending the choice of a final mode of management (e.g. underground disposal). The assay system, called UTM (the French acronym for “Measurement and Sorting Unit”), consists of three stations devoted to active gamma imaging, gamma-ray spectroscopy and combined passive / active neutron measurements. After nearly 3 years of optimization and design studies [1], the CEA has built a scale one mock-up of UTM, called SYMETRIC. The purpose was to validate the performances formerly assessed by numerical simulation, mainly with the computer code MCNP [2]. We present here the experimental results obtained with SYMETRIC for five real bituminized waste drums. These confirm the expected performances in the measurement time assigned for each assay, which is limited to 1200 seconds. With the help of gamma imaging, we are able to determine the density of the bituminous mix with an uncertainty of ± 10% for a confidence level of 95%. We can also measure the filling height with an accuracy of ± 2 cm. These data allow us to correct matrix effects in gamma and neutron measurements. For these assays, the main results concern the detection limits and measurement uncertainties on 241 Am, 239 Pu and 240 Pu. These radioisotopes represent the major part of the total alpha activity, which is a very sensitive parameter for surface disposal limited to a maximum level of about 10 GBq per drum. The alpha activity must be calculated after a radioactive decay of 300 years, which is the survey period of the French Aube Center. If we can detect the former isotopes, the uncertainties on their measured activities are roughly 50%. If not, the detection limits are around a few GBq. These performances are sufficient to allow the sorting of the drums to either surface repository or interim storage. However, in order to increase the margin between the detection limits and the acceptance criterion on the total alpha activity, additional studies on the optimization of the measurement performances will be carried out. In this context, the experience gained with the SYMETRIC mock-up will be very helpful.Copyright

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.

and

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}

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing [1]. 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 137Cs and 134Cs 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 137Cs 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 3He 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 stations.

and From Photofission of

The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA NC La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT). In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (LMN) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor, namely a 3He proportional counter located inside the measurement cavity. After feasibility studies performed with LMNs PROMETHEE 6 laboratory measurement cell and with MCNPX simulations, this paper presents first experimental tests performed on the industrial ACC (hulls and nozzles compaction facility) measurement system. A calculation vs. experiment benchmark has been carried out by performing dedicated calibration measurements with a representative drum and 235U samples. The comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach and the industrial feasibility of the method, which will be implemented on the industrial station for the measurement of historical wastes.

^{235} {\rm U}

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.