Gamma Heating Measurements Research Articles

Validation and improvement of gamma heating calculation methods for the G.A. Siwabessy multipurpose reactor

Gamma heating, which is deposited in irradiated targets or samples, is an important issue in research reactors because it affects the safety of samples and the reactor operation. Gamma heating in the Reaktor Serba Guna G.A. Siwabessy (RSG GAS) can be measured or calculated using computer code. In this paper, we present the results obtained by using gamma calorimeters to measure the gamma heating in the central irradiation position in RSG GAS. The measurement results were verified and then compared with the calculation results obtained using the Gamset code. However, the accuracy of the calculation results obtained using Gamset was inadequate, with more than 20% error. Moreover, Gamset was initially created to calculate gamma heating in 35 MWth reactors, and the RSG GAS reactor has an operating power level of 30 MWth. To address these issues, we developed a new program called NewGamset, built with an informative graphical user interface-based display. NewGamset can employ new analytical approaches as it utilizes a calculation base that comprises 18 energy groups and can provide updated physical parameters of RSG GAS. The results of gamma heating measurements obtained using gamma calorimeters made of graphite, aluminum, iron, and zirconium were 2.20 ± 0.03 W/g, 2.25 ± 0.02 W/g, 2.58 ± 0.03 W/g, and 2.91 ± 0.10 W/g, respectively. In general, the calculation results were larger than the measurement results, and the average ratio of calculation to measurement result was 1.37 for Gamset and 1.02 for NewGamset. Hence, it can be concluded that NewGamset provides more accurate results than Gamset. We also report the gamma heating estimation for common elements irradiated in the RSG GAS silicide core, with an operating power level of 15 MWth and 30 MWth. We highly recommend that NewGamset be used for validating the gamma heating data of various target materials in RSG GAS.

Computational support on the development of nuclear heating calorimeter detector design

Heating due to energy deposition of intense ionizing radiation in samples and structural materials of nuclear reactors poses severe limitations in terms of cooling requirements for safe reactor operation, especially in high neutron and gamma flux environments of material testing fission reactors (MTRs) and novel fusion devices. A bilateral CEA-JSI research project was launched in 2018 with the objective to measure the gamma heating rates in standard reactor-related materials (graphite, aluminium, stainless steel and tungsten) as well as fusionrelevant materials (low-activation steel Eurofer-97 and Nb3Sn superconductor) in the JSI TRIGA reactor my means of gamma calorimeters. The calorimeter design will be based on the the CALMOS-2 calorimeter developed at the CEA and used to perform gamma heating measurements in the OSIRIS MTR in Saclay. In order to optimize the detector response inside the JSI TRIGA reactor field and not to perturb the measurement field, a detailed computational analysis was performed in terms of energy deposition assessment and measurement field perturbation using the MCNP v6.1 code, and in terms of heat transfer using the COMSOL Multiphysics code. The abovementioned activities enabled us to finalize the detector design with the experimental campaign planned for the end of year 2019.

Tests and Foreseen Developments of Fibered-OSLD Gamma-Heating Measurements in Low-Power Reactors

In this paper, test measurements of a fibered optically stimulated luminescent dosimeter (OSLD) system performed during a dedicated experimental phase in EOLE zero-power reactor are presented. The measurement setup consists of an OSLD crystal connected onto the extremity of an optical fiber and a laser stimulation system, manufactured by the CEA/LIST, Saclay, France. The OSL sensor is remotely stimulated via an optical fiber using a diode-pumped solid-state laser. The OSL light is collected and guided back along the same fiber to a photomultiplier tube through appropriate optical filters. OSL signals are corrected for background noise and integrated to provide the absorbed doses thanks to calibration coefficients. The light stimulation also fully bleaches the crystal (reusable for the next measurement). Results obtained using this system are compared to usual gamma-heating measurement protocol using OSLD pellets. Both measurement processes were calibrated to ensure the comparison of absolute doses in gray at the same positions in the core. The presence of induced radioluminescence in the OSLD during the irradiation was also observed and could be used to monitor the gamma flux. The feasibility of remote measurements is achieved, whereas further developments could be conducted to improve this technique since the readout procedure still requires to withdraw the OSLD off the gamma flux (hence from the core) on account of the dose rate (around a few Gy $\cdot \text{h}^{\mathrm {\mathbf {-1}}}$ ), and the readout time remains quite long for online applications. Several improvements are foreseen and will be tested in the forthcoming years.

The CANDELLE Experiment for Characterization of Neutron Sensitivity of LiF TLDs

As part of the design studies of future power and research nuclear reactors, the validation of neutron and photon calculation schemes requires the implementation of nuclear heating measurements. Such measurements are usually performed in very low-power reactors, whose core dimensions are accurately known and where irradiation conditions are entirely controlled. The use of thermoluminescent dosimeters is particularly well suited to gamma heating measurements, although the TLDs are also sensitive to neutrons depending on their composition. Hence, for gamma heating assessment in a mixed gamma–neutron field, the neutron sensitivity correction factors of TLDs has to be accurately determined to ensure a better interpretation of doses measured in a reactor with reduced uncertainties. This paper presents the CANDELLE experiment, which is dedicated to the implementation of a pure neutron field calibration of lithium-enriched TLDs, by using the Platform for Studies and Research on Nuclear Energy platform of LPSC. This calibration step is part of a promising approach that has been recently developed at INFN for medical purposes, aiming at determining the gamma and neutron contributions to the doses measured by LiF TLDs in the mixed field, from the peak height ratios of the gamma and neutron calibration glow curves of this type of TLDs. The test irradiations conducted at LPSC within the first phase of the CANDELLE project allowed retrieving the peak height of LiF TLDs in fast neutron field, hence providing an early assessment of the neutron components of doses measured in the EOLE reactor at CEA Cadarache with around 10% uncertainty at $1\sigma$ . From these preliminary results, the second phase of the project will be devoted, on the one hand, to the use of moderator screens aiming at achieving a thermal neutron field while avoiding a too strong attenuation of the flux, and on the other hand, to the determination of the neutron spectrum at different locations within the irradiation area.

Tests and foreseen developments of fibered-OSLD gamma heating measurements in low-power reactors

In this paper are presented test measurements of a fibered-OSLD system performed during a dedicated experimental phase in EOLE zero-power reactor. The measurement setup consists of an OSLD crystal connected onto the extremity of an optical fiber and a laser stimulation system, manufactured by the CEA/LIST in Saclay. The OSL sensor is remotely stimulated via an optical fiber using a diode-pumped solid-state laser. The OSL light is collected and guided back along the same fiber to a photomultiplier tube. Results obtained using this system are compared to usual gamma heating measurement protocol using OSLD pellets. The presence of induced radio-luminescence in the OSLD during the irradiation was also observed and could be used to monitor the gamma flux. The feasibility of remote measurements is achieved, whereas further developments could be conducted to improve this technique since the readout procedure still requires to withdraw the OSLD off the gamma flux (hence from the core) on account of the dose rate (around a few Gy.h-1), and the readout time remains quite long for on-line applications. Several improvements are foreseen, and will be tested in the forthcoming years.

Analysis and Recent Advances in Gamma Heating Measurements in MINERVE Facility by Using TLD and OSLD Techniques

Analysis and Recent Advances in Gamma Heating Measurements in MINERVE Facility by Using TLD and OSLD Techniques

CALMOS: Innovative Device for the Measurement of Nuclear Heating in Material Testing Reactors

Abstract An R&D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA’s Nuclear Energy Division at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2 W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed.

Gamma Heating Measurements in a Mixed Radiation Field

Gamma hearing measurements have been made in a low-Z assembly irradiated with 14-MeV neutrons and (n,n{prime}) gammas produced by a Texas Nuclear Model 9400 neutron generator. The assembly is composed of 144 magnesium sleeves (5cm {times} 5cm {times} 60cm {times} 3 mm thick) filled with graphite to simulate a fusion blanket test module. Heating measurements were made in the mid-line of the assembly using a proportional counter operating in the Continuously-varied Bias-voltage Acquisition (CBA) mode. The neutron induced atomic recoil signal was rejected by observing the signal rise-time differences inherent to radiations of different LET. The experiment was modelled using the one-dimensional radiation transport code ANISN/PC. The operating limits of this technique were identified by comparing measurements made at different positions in the assembly and then comparing these measurements to the calculated flux. 7 refs., 5 figs., 1 tab.

Gamma heating measurements with proportional counters

A new data acquisition technique (the continuously-varied bias-voltage acquisition mode) has been developed and tested for the low-flux broad-energy regime characteristic of existing fusion blanket mockups. This method of analysis allows for the acquisition of data spanning several orders of magnitude in energy with a single proportional counter. Utilizing this method, the gamma energy deposition in a mixed neutron and gamma field was measured.