Fission Chamber Measurements Research Articles

Performance Assessment of Amplification and Discrimination Electronic Devices for Passive Neutron Measurements

The assessment of fissile material mass is a key requirement to enhance radioactive waste management and to ensure a high level of safety in the nuclear industry. The assessment of plutonium fissile mass by passive coincidence measurements is usually obtained by detecting the neutrons generated by the spontaneous fissions of Pu isotopes. The detection and the quantification of a fissile mass in the radioactive waste can also be carried out using active neutron interrogation with a pulsed D-T neutron generator. The 14-MeV neutrons are moderated to induce fissions in the uranium and plutonium-contaminated waste. The emitted neutron signal can be analyzed by considering their detection time after the generator pulse. Data are measured according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and good gamma-ray immunity, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas, and a threshold discriminator produces a logic pulse used for neutron counting. This article describes the performance assessment of different commercially-available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to the Mirion 7820 current amplifier used in specific high-count rate applications. An optimization of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.

Performance assessment of amplification and discrimination electronic devices for passive neutron measurements

The knowledge of the fissile material mass is a key challenge to enhance radioactive waste management and to ensure a high level of safety in nuclear industry. Data is analyzed according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and a good gamma-ray discrimination, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas and a threshold discriminator produces a logic pulse used for neutron counting. This paper describes the performance assessment of different commercially available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to Mirion 7820 current amplifier used in specific high-count rate applications. An optimisation of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.

Investigation of fission chamber response in the frame of fuel debris localization measurements at Fukushima Daiichi

This work aims at assessing the performance of a235U enriched fission chamber in order to localize fuel debris, prior to dismantling operations, in a flooded primary containment vessel of a damaged nuclear reactor such as Fukushima Daiichi. Based on both a comprehensive scan of the environment and the detection of neutrons emitted by the melted core, fuel debris can be localized. In this paper, we carry out a simulation study using the MCNP6 code to investigate fission chamber response in the frame of fuel debris localization measurements in a damaged nuclear reactor. The CFUF34 fission chamber (manufactured by PHOTONIS) and the primary containment vessel of Fukushima Daiichi Unit 1 were chosen to conduct this work. Impact of different parameters were investigated with MCNP6, such as: neutron energy, water temperature, fission chamber position (altitude, lateral shift, and rotation), and sensitivity loss due to sediments potentially covering fuel debris. In summary, we show that fuel debris should be sought by their thermal neutron signature at a distance of a few centimeters and that potential rotational movements of the fission chamber up to 60° have a limited impact on signals measured. We also show that sensitivity loss due to sediments potentially covering fuel debris has been evaluated on the order of a factor 10 considering a 30 cm-thick sediment layer. On the other hand, experiments were performed to assess the impact of a strong gamma dose rate on fission chamber measurements. These irradiation trials involved a CFUE32 fission chamber (also manufactured by PHOTONIS) available in our laboratory and three different irradiation means: an X-ray tube, an 192Ir source, and a linear electron accelerator. These experiments enable to draw the conclusion that the fission chamber is not impacted by the gamma dose rate up to 104 Gy h−1, which is in good agreement with specifications provided by the manufacturer (PHOTONIS). In addition, no performance degradation was observed after an integrated gamma dose of 2200 Gy on the fission chamber in a 10 min irradiation. However, when the fission chamber is irradiated by gamma dose rates above 104 Gy h−1 (upper limit of the operating domain specified by PHOTONIS), a significant gamma background is observed. Nevertheless, as the gamma dose rates at Fukushima Daiichi should not exceed 103 Gy h−1, fission chamber measurements performed towards fuel debris localization in the primary containment vessels of the units would not be affected by the severe gamma-ray irradiation.

A Re-Analysis of Historical Los Alamos Critical Assembly Reaction Rate Measurements

Starting in the 1950s and continuing into the early 1970s, a number of foil irradiations and fission chamber measurements were made in a variety of Fast critical assemblies at Los Alamos National Laboratory. These include (i) Godiva, a bare HEU spherical assembly; (ii) Flattop-25, a spherical assembly consisting of an HEU core and a natural uranium reflector; (iii) Jezebel, a bare 239 Pu assembly; and (iv) Flattop-Pu, a spherical assembly consisting of a 239 Pu core and a natural uranium reflector. In most instances the irradiations occur at or near the center of the assembly, but in selected instances data were obtained for a radial traverse extending into the Flattop reflector region. Measurements were made for a number of threshold reactions, including 45 Sc(n,2n)44m Sc, 51 V(n ,α)48 Sc, 75 As(n,2n)74 As, 89 Y(n,2n)88 Y, 90 Zr(n,2n)89 Zr, 103 Rh(n,2n)102g Rh, 107 Ag(n,2n)106m Ag, 169 Tm(n,2n)168 Tm, 175 Lu(n,2n)174 Lu, 191 Ir(n,2n)190 Ir, 197 Au(n,2n)196 Au, 203 Tl(n,2n)202 Tl, 204 Pb(n,2n)203 Pb and 238 U(n,2n)237 U. Fission ratio data for 238 U(n,f)/235 U(n,f) and 239 Pu(n,f)/235 U(n,f) were also obtained. We report C/E values from MCNP6 calculations using ENDF/B-VII.1 and IRDFF-v1.03 cross section data.

A Fission Chamber Measurement of the natB(d,n) Cross Section for Use in Neutron Detector Calibration

The use of an easily reproducible neutron source reaction that produces a well-known continuous spectrum of neutrons over a range of energies is an ideal solution for some neutron detector efficiency calibrations. Fission chamber measurements of the 27Al(d,n) reaction have proven valuable for detector calibration for energies between 0.2 and 14 MeV. To complement the aluminum data, measurements were made with a fission chamber at 60 deg of the neutron spectrum produced from the 7.5-MeV deuteron bombardment of a thick natural boron target. This should enable accurate and efficient calibration of neutron detectors for the energy range between 0.09 and 19.6 MeV. Tenth-order polynomial fits to the data are provided for the region with energies between 88 keV and 2.33 MeV and the region with energies between 1.76 and 19.6 MeV.

Comparative Study of Measurements by Means of Gamma Thermometer Strings with Fission Chamber Measurements

A comparison of local linear power densities measured with a gamma thermometer string (GTS) and with a fission chamber shows important discrepancies at the extremities of the assembly. The cause wa...