Small Fission Chambers Research Articles

Calibration of Large Photonis Fission Chambers in Standard Neutron Fields of the BR1 Reactor

Small fission chambers manufactured by Photonis, Reuter-Stokes, CEA, and Centronic are routinely calibrated in standard neutron fields of the BR1 reactor at SCK CEN in Mol, Belgium. Two irradiation fields are available: fast spectrum in the MARK3 convertor and thermal spectrum in the Empty Cavity. In this work, for the first time the calibration procedure of larger fission chambers (Photonis CFUL01 and CFUM21 type) with the deposit length exceeding the length of the convertor is presented. Spatial energy correction factors taking into account the neutron flux shape beyond the convertor and radial gradient of the neutron flux were calculated with MCNP and experimentally validated. The combination of calibration (i.e., effective mass measurement) in fast and thermal irradiation fields allows for determination of the purity of ²³⁸U deposit. This is demonstrated on three CFUL01-type fission chambers with purity between 99.8% and 99.998%.

Small Thorium Fission Chambers for Recording Fast Neutrons in Reactor Installations

The results of development work on experimental samples of small fission ionization chambers for recording fast neutrons (En ~ 1–20 MeV) in a reactor core are presented. A technological solution for obtaining a radiator based on ThO2 as a result of nonstationary low-temperature consolidation of an oxide coating from a solution of thorium carboxylate was developed and realized. Samples of ThO2-based radiators on the surfaces of multichannel electrodes of diameter 3 and 5 mm and coatings based on the isotope 238U (in the form U3O8 based on uranium with 235U content 0.00002%) were fabricated and small fission chambers based on them were assembled. Sample radiators were bench tested for stability during thermal cycling (15 cycles, heating to 750°C) and under vibration loads (to 200 Hz). Physico-mechanical and reactor tests of the fission chambers were conducted and the discrimination characteristics of samples and their sensitivity to heat and fast-neutron fluxes were determined. A comparative analysis of the conditions of effective application of small 238U- and 232Th-based fission chambers in long-time operation as well as in single neutron measurements is made.

On Use of Multi-Chambered Fission Detectors for In-Core, Neutron Spectroscopy

Presented is a short, computational study on the potential use of multichambered fission detectors for in-core, neutron spectroscopy. Motivated by the development of very small fission chambers at CEA in France and at Kansas State University in the U.S., it was assumed in this preliminary analysis that devices can be made small enough to avoid flux perturbations and that uncertainties related to measurements can be ignored. It was hypothesized that a sufficient number of chambers with unique reactants can act as a real-time, foilactivation experiment. An unfolding scheme based on maximizing (Shannon) entropy was used to produce a flux spectrum from detector signals that requires no prior information. To test the method, integral, detector responses were generated for singleisotope detectors of various Th, U, Np, Pu, Am, and Cs isotopes using a simplified, pressurized-water reactor spectrum and fluxweighted, microscopic, fission cross sections, in the WIMS-69 multigroup format. An unfolded spectrum was found from subsets of these responses that had a maximum entropy while reproducing the responses considered and summing to one (that is, they were normalized). Several nuclide subsets were studied, and, as expected, the results indicate inclusion of more nuclides leads to better spectra but with diminishing improvements, with the best-case spectrum having an average, relative, group-wise error of approximately 51%. Furthermore, spectra found from minimum-norm and Tihkonov-regularization inversion were of lower quality than the maximum entropy solutions. Finally, the addition of thermal-neutron filters (here, Cd and Gd) provided substantial improvement over unshielded responses alone. The results, as a whole, suggest that in-core, neutron spectroscopy is at least marginally feasible.

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

During the GUINEVERE FP6 European project (2006–2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2 O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

New Interpretation of Signals and In-Reactor Detector Calibration

One way to increase the accuracy of monitoring general and distributed parameters of the core of a nuclear reactor is to calibrate the in-reactor neutron and γ-ray detectors. The calibration can be performed before the detectors are placed in the core as well as periodically during operation. The approaches used to calibrate detectors for different types of modern nuclear reactors are examined and the accuracy of some methods of calibration is presented. Calibration of in-reactor neutron and γ-ray detectors is needed in order to secure acceptable accuracy in monitoring general and distributed parameters of a reactor core, including the margin to the safe operating limit. It is well known that more accurate monitoring of the distributed parameters makes it possible to increase the average energy density of the fuel in the region of the ‘plateau’ in the power release distribution, decrease neutron leakage from the core and thereby increase fuel burnup [1]. The calibration can be done before the detectors are placed into the reactor as well as periodically in the course of operation. PWR, BWR. Movable in-reactor probes, such as small fission chambers, equipped with flexible communication lines and special drives, are the most commonly used means for performing periodic calibration of in-reactor detectors for PWR and BWR reactors. The chambers are moved in dry channels, passing through the bottom of the reactor vessel [2]. The number of probes ranges from 2 to 5, and all of them can scan in turn each of 43‐64 sleeves. KWU. A pneumatic ball system is used to scan the neutron flux density in the German KWU reactors. The system activates balls containing vanadium in 28 dry sleeves placed in the core. The balls are put in and removed under gas pressure. The activity of the balls is calculated outside the reactor vessel [2]. The neutron flux density in each stationary detector (fission chamber or emission detector) is reconstructed according to the indications of the probes or activity of the balls and the sensitivity is determined for the measurement time. The scan results can also be used to periodically monitor the error in the reconstruction of the neutron flux density distribution in the core. g-Sensitive Thermocouples for BWR. In BWR vessel reactors, thermocouples measuring the temperature differential between the sensitive element (heated by γ-radiation from the core) and the detector shell (cooled by the coolant) are being used once again to extend the service life of in-reactor detectors. To measure the γ-radiation dose rate, it is necessary to know its relation to the temperature difference, which depends on the thermal conductivity of the material (the thermal conductivity of the material changes as the fluence builds up) separating the heated elements and the detector casing.

Application of Monte Carlo Method to Solve the Neutron Kinetics Equation for a Subcritical Assembly

There is a need to understand the space-dependent kinetics of fast or thermal reactor physics and the Monte Carlo method should be implemented in kinetics codes as well. In a transient accident (for example, control rod ejection accident or loss of coolant accident), changes in the system are much slower than the prompt neutron lifetime. In the present paper, a method is proposed in which the Monte Carlo method is used to solve the neutron kinetics of reactors, such as the case of a subcritical static reactor with an external neutron source. In the proposed method the time derivative of the neutron equation is set to zero, though the time derivative of the delayed neutron precursor equation is treated without any approximation in the time differentiation of delayed neutron precursor equation. Most kinetics calculation methods originate from the point kinetics method, and they require calculation of excess reactivity at each time step, but the present method does not need calculation of excess reactivity. The neutron counts measured with small fission chambers in the subcritical kinetics experiment using the JAEA fast critical assembly are compared with counts calculated by the proposed method and compared with the measured values. The agreement is good, and the excellent performance of the method for space-dependent neutron kinetics is shown.

Use of thermal-neutron probes to measure thermal-neutron flux of distributions

At the present time considerable attention is being paid to the monitoring of energy release in the cores of nuclear reactors for various purposes [i]. Exact and reliable measurement of the neutron fields in reactor cores requires continuous improvement of old experimental methods and development of new ones. In-reactor measurements, however, always entail certain difficulties due to the difficult radiation and thermal conditions under which the detectors must operate, etc. In recent years there has been a great development of instrumental methods of measuring neutron fields, methods in which small fission chambers, electron-emission neutron, self'powered probes, thermal-neutron probes, etc. are used as the primary probes [2-5].

Development and Testing of Neutron Dosimetry Techniques for Accelerator-Based Irradiation Facilities

Integral tests have been made for 27 nuclear activation cross sections required for neutron dosimetry at accelerator-based irradiation facilities. Thirteen materials were irradiated in a well-defined geometry using neutrons produced by stopping 14- to 16-MeV deuterons in a thick beryllium target. The neutron spectrum was measured by time-of-flight spectrometry. Comparison of the activation integrals with the differential neutron spectrum provided an assessment of the nuclear data. A small fission chamber was used to monitor the irradiation, map flux gradients, and study attenuation and scattering effects. Comparison of the fission rates to fission product activations in the foil materials was also used to measure cumulative fission yields for /sup 235/U and /sup 238/U.

The behavior of ionization chamber-neutron detectors in intense transient gamma fields

The use of fast neutron detectors that are only slightly sensitive to gamma radiation coupled with fast preamplifiers and a linear gate has created the capability of measuring the time behavior of the neutron population very close in time to the bremsstrahlung burst from a linear accelerator. Three neutron detectors were investigated in high transient gamma fields. They are: totally depleted silicon solid state detector with fission conversion foil, small fission chamber and small 10B-coated proportional counter. These detectors coupled with proper electronics have yielded reliable results as early as few hundred nanoseconds after the gamma burst.

Measurements of buckling and relative reaction rates in some plutonium-graphite assemblies

The materials bucklings of a series of plutonium-aluminium alloy-fuelled, graphite-moderated assemblies have been measured, the carbon/plutonium ratio ranging from 14,520 to 2420. Fission ratios of 239Pu, 235U and 233U in the assemblies have been obtained using small fission chambers. On one of the systems, experiments were extended up to a temperature of 370°C. The experimental data have been compared with the predictions of a 43-group diffusion-theory code. Agreement between predicted and measured bucklings is good over the whole range of composition, within the limits of experimental error, which are equivalent to an uncertainty of about ± 2 per cent in reactivity. The spectrum-sensitive 239 Pu 235 U fission ratios are 6–9 per cent higher than those calculated from the multigroup spectra. Comparison with further results from an assembly loaded with 235U in similar configuration, leads to a ratio of the 2200 m/sec η-values of 239Pu and 231U of 1·02 ± 0·02.