Measuring gamma doses over the mGy-kGy range with a single type of TLD detector

Thermo-luminescent detectors are currently used to measure gamma doses in the medical and environmental surveillance fields. During the past few years, the CEA Reactor Studies Division tested and validated the use of these detectors for gamma flux characterization and nuclear heating measurements in mixed neutron/gamma fields of low power reactors. Doses were comprised between a few mGy and a few Gy for dose rates up to a few Gy.h-1. However, in MTR or TRIGA reactors, the gamma flux level is much higher (> 1012 n/cm2/s) and the TLD currently in use (CaF2:Mn and 7LiF:Mg,Ti) and their readout protocols were no longer suitable for the resulting doses. In order to extend the applicable dose range up to ∼1 MGy (dose rate of a few kGy.h-1), several options were explored. On one side, some adjustments were made to the readout protocols of CaF2:Mn and 7LiF:Mg,Ti, notably by testing the use of filters to reduce the amount of light received by the reader PMT to avoid saturation. On the other side, a new type of TLD (LiF:Mg,Cu,P) with different Li enrichments (natural or enriched in 7Li) was tested. This paper presents the calibration measurements results performed in pure gamma fields, at the irradiation platform of the CEA Cadarache Radioprotection Division, between 250 mGy and 3 Gy for all detector types. In addition to the calibration, these measurements also studied the Mg,Cu,P-doped detectors response: reproducibility, dose rate dependence, incoming photon energy dependence, high temperature effect when reading TLD, etc. Results show that at low doses Mg,Cu,P-doped TLDs are slightly less stable than CaF2:Mn and 7LiF:Mg,Ti. The sensitivity modification after a high dose exposure seems to indicate that a new protocol readout should be defined for Mg,Cu,P-doped sensors (high temperature peak).

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.

Characterization of the gamma flux in a tangential channel of the CENM TRIGA MARK II research reactor

The CNESTEN (National Center for Energy Sciences and Nuclear Technology, Morocco) operates a TRIGA Mark II reactor, which can reach a thermal maximum power at steady state of 2 MW. In reactors devoted to research and experiments, it is mandatory to characterize the neutron and photon fields in the irradiation positions. Together with a computational model of the core, it ensures the ability to reach the requested uncertainties when performing experiments, such as detectors testing, irradiation for hardening or nuclear data measurements. The neutron field of different irradiation positions has been characterized by dosimetry techniques and compared to the MCNP full model of the reactor. Preliminary photon propagation calculations are also performed with this model, but up to now, no experimental validation of the results exists. The aim of the newly set collaboration between CEA and CNESTEN is to characterize the gamma field of these positions. The first position investigated is the part of the NB1 tangential channel closest to the core. Among gamma measurements techniques, and according to the constraints arising from using this channel, it was chosen to use thermos- and optically stimulated luminescent detectors. This paper presents the experiments carried out in September 2018 as well as their results. Three detectors types were used: TLD400 (CaF2:Mn), TLD700 (7LiF:Mg,Ti) and OSLD (Al2O3:C). Measurements were performed in several steps: background measurements, transient measurements (divergence phase + SCRAM), and irradiation at steady state. In the end, these measurements will provide a dose as well as a gamma flux value for this position.

Radiation hardness studies and detector characterisation at the JSI TRIGA reactor

The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.

Gamma-heating and gamma flux measurements in the JSI TRIGA reactor, results and prospects

The neutron field of various irradiation positions of the TRIGA Mark II reactor of the Jožef Stefan Institute has been thoroughly characterized by neutron activation dosimetry and miniature fission chambers techniques. In order to have a fully validated calculation scheme to analyze and plan experiments, the gamma field also has to be experimentally validated. The 10-year long collaboration between CEA and JSI is a perfect framework to carry out such a study, and measurements of the gamma field started in late 2016. Several measurement techniques were investigated in in-core and ex-core positions. On-line measurements were carried out using miniature ionization chambers manufactured by the CEA and PTW Farmer ionization chambers. Positional dependence was studied, showing a decrease in the delayed gamma contribution to the total gamma flux with increasing distance from the reactor core center. To characterize the gamma dose in the core, as well as in the periphery, thermo- and optically stimulated luminescent detectors were tested. These detectors are commonly used at CEA to measure the gamma dose in a given material in order to study the nuclear heating in various core elements (control rod, baffle, structural material). Different filters were used in order to assess an integrated dose ranging from a few Gy up to several kGy. The feasibility of such measurements demonstrates the complementarity between measurements with dosimetry and ionization chambers from low to very high gamma-dose environment, such as in material testing reactors.

Gamma-induced background in the KATRIN main spectrometer

The KATRIN experiment aims to measure the effective electron antineutrino mass $$m_{\overline{\nu }_e}$$mν¯e with a sensitivity of $${0.2}\,{\hbox {eV}/\hbox {c}^2}$$0.2eV/c2 using a gaseous tritium source combined with the MAC-E filter technique. A low background rate is crucial to achieving the proposed sensitivity, and dedicated measurements have been performed to study possible sources of background electrons. In this work, we test the hypothesis that gamma radiation from external radioactive sources significantly increases the rate of background events created in the main spectrometer (MS) and observed in the focal-plane detector. Using detailed simulations of the gamma flux in the experimental hall, combined with a series of experimental tests that artificially increased or decreased the local gamma flux to the MS, we set an upper limit of $${0.006}\,{\hbox {count}/\hbox {s}}$$0.006count/s (90% C.L.) from this mechanism. Our results indicate the effectiveness of the electrostatic and magnetic shielding used to block secondary electrons emitted from the inner surface of the MS.

NEUTRON AND GAMMA SPECTRUM ANALYSIS OF KARTINI RESEARCH REACTOR FOR BORON NEUTRON CAPTURE THERAPY (BNCT)

MCNPX was used to design a three-dimensional model of Kartini Research Reactor (KRR) as a neutron source and performed criticality calculation. The criticality calculation of the reactor aims to obtain the neutron and gamma spectrum by simulating the fission reaction inside the reactor core. Total source histories were 105 per cycle, when the number of cycle for criticality calcutation was 1000 cycles with 60 skipped cycles. The reactor criticality according to the simulation result is 1.00179±0.00007. The total neutron flux on ring A, B, C, D, E and F inside the reactor core are respectively 6.553×1011 n/cm2s, 4.53×1012 n/cm2s, 4.167×1012 n/cm2s, 3.751×1012 n/cm2s, 2.914×1012 n/cm2s and 3.107×1012 n/cm2s. The total gamma flux is 6.956×1011 particles/cm2s, 4.838×1012 particles/cm2s, 4.398×1012 particles/cm2s, 3.962×1012 particles/cm2s, 2.953×1012 particles/cm2s and 2.013×1012 particles/cm2s, respectively for each ring. Thermal neutron fluxes recorded on the base of radial piercing beamport were 4.678×1010 n/cm2s, with the epithermal neutron flux of 5.37×109 n/cm2s and fast neutron flux of 4.17×1010 n/cm2s. The gamma flux on that side reaches 4.22×1012 particles/cm2s. On the 92-cm-ranges from the base inside radial piercing beamport, both neutron and gamma flux decrease up to 5.11×108 n/cm2s for thermal neutron flux, 4.598×106 n/cm2s for epithermal neutron flux, 2.55×107 n/cm2s for fast neutron flux and 8.214×1010 particles/cm2s for gamma flux. In conclusion, the spectrum yield from this study can be use to define the source spectrum of the simulations and optimations prior to BNCT pre-clinical trial (in vivo/in vitro test) use KRR radial piercing beamport.Keywords: BNCT, radial piercing beamport, Kartini Research Reactor, neutron spectrum, gamma spectrum ANALISIS SPEKTRUM NEUTRON DAN GAMMA UNTUK BORON NEUTRON CAPTURE THERAPY (BNCT) DI REAKTOR KARTINI. MCNPX telah digunakan untuk memodelkan bentuk 3 dimensi dari Reaktor Kartini sebagai sumber neutron dan melakukan perhitungan kekritisan. Perhitungan kekritisan reaktor bertujuan untuk mendapatkan spektrum neutron dan gamma dengan mensimulasikan reaksi fisi yang terjadi di dalam inti reaktor. Jumlah source histories adalah 105 per iterasi, dimana banyaknya iterasi yang dilakukan dalam perhitungan kritikalisasi adalah 1000 iterasi dengan jumlah iterasi yang dilewatkan adalah 60 iterasi. Nilai kekritisan reaktor sesuai dengan hasil simulasi adalah 1,00179±0,00007. Fluks neutron total pada ring A, B, C, D, E and F di dalam inti reaktor masing-masing adalah 6,553×1011 n/cm2s, 4,53×1012 n/cm2s, 4,167×1012 n/cm2s, 3,751×1012 n/cm2s, 2,914×1012 n/cm2s and 3,107×1012 n/cm2s. Total fluks gamma adalah 6,956×1011 partikel/cm2s, 4,838×1012 partikel/cm2s, 4,398×1012 partikel/cm2s, 3,962×1012 partikel/cm2s, 2,953×1012 partikel/cm2s dan 2,013×1012 partikel/cm2s, masing-masing untuk tiap ring. Fluks neutron termal hasil perekaman pada pangkal beamport tembus radial adalah 4,678×1010 n/cm2s, dengan fluks neutron epitermal sebesar 5,37×109 n/cm2s dan fluks neutron cepat sebesar of 4,17×1010 n/cm2s. Fluks gamma pada bagian tersebut mencapai 4,22×1012 partikel/cm2s. pada jarak 92 cm dari pangkal beamport tembus radial, fluks neutron dan gamma turun mencapai 5,11×108 n/cm2s untuk fluks neutron termal, 4,598×106 n/cm2s untuk fluks neutron epitermal, 2,55×107 n/cm2s untuk fluks neutron cepat dan 8,214×1010 partikel/cm2s untuk fluks gamma. Kesimpulannya, spektrum yang dihasilkan pada studi kali ini dapat digunakan untuk mendefinisikan sumber dalam simulasi dan optimasi terutama untuk keperluan uji pre-klinis (uji in vivo/ in vitro) BNCT menggunakan beamport tembus radial Reaktor Kartini. Kata kunci: BNCT, beamport tembus radial, Reaktor Kartini, spektrum neutron, spektrum gamma