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Nuclear isomers are investigated to perform precise epithermal neutron dosimetry. One key physical property for reactor dosimetry is the precise knowledge of the isomer half-life. In the list of potential candidates, the half-life values available in the literature or in the recommended database for reactor dosimetry for 117mSn, 125mTe, 135mBa and 195mPt show some discrepancy. New half-life measurements are presented in this work. Those isomers were produced by neutron activation at the JSI TRIGA Reactor from tin, tellurium, barium (all three of them enriched in their parent of interest) and natural platinum dosimeters. Precise half-lives measurements were performed at the MADERE platform using gamma and X-ray spectrometry. Special care has been taken on the uncertainty determinations. • New elements for fast neutron dosimetry are currently under investigation. • Metastable states of Tellurium, Tin, Barium, and Platinum are promising candidates. • Half-lives are key elements in reactor dosimetry. • For nuclear data assessment, uncertainties must be determined with high accuracy. • Good agreement of the measurement with the litterature. [ABSTRACT FROM AUTHOR]

The Jordan Research and Training Reactor (JRTR) is a multi-purpose reactor with power of 5 MW currently under commissioning. The core consists of standard MTR plate type fuel assemblies with low enriched fuel of 19.75% U-235 enrichment. A new computational model has been developed for JRTR using the MCNP6 Monte Carlo code. The purpose of this paper is to validate the MCNP6 model by comparison of the calculation results of important neutronics parameters like k eff , flux distribution, kinetics parameters, power peaking factors, and control rod worth to the reference results which were obtained by the McCARD Monte Carlo code and presented in the safety analysis report. Same modeling assumptions were adopted in the MCNP6 model in order to check the computational differences. In addition various up-to-date nuclear data libraries were used in calculations to assess their effect on calculated quantities. The calculation results showed good agreement and the difference in the effective multiplication factor is up 50 pcm. [ABSTRACT FROM AUTHOR]

Abstract: The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the 197Au(n,γ)198Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. [Copyright &y& Elsevier]

CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. [ABSTRACT FROM AUTHOR]

Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate neutron flux evaluation, well-known and reliable cross sections are needed. In the thermal and fast energy region, many reliable monitoring reactions exists, however, in case of the epithermal and intermediate energy region, there are practically no dosimetry nuclear reactions sensitive specifically in this energy range. Due to this fact, both new data are being measured and methodologies are under development to describe and test this energy region. It was found that various neutron filters can be used to cut parts of neutron spectra and thus methodology based on spectrum filtering could potentially be employed to survey cross sections of interest. It this paper, the use of 3 different filters – B 4 C, Cd, and In is studied, on the case of the 55Mn(n,γ) reaction. Measured values of that cross section in the given filtered reference spectra are reported. • Measurement of reaction rates in various neutron filters in reference neutron field. • Calculation of reaction rates in various neutron filters in reference neutron field. • Discrepancies between calculation and experiment. • Testing of 55Mn(n,γ) cross section in various energy regions. [ABSTRACT FROM AUTHOR]

ABSTRACT: Ionizing radiation has been identified as an option for sterilization of disposable filtering facepiece respirators in situations where the production of the respirators cannot keep up with demand. Gamma radiation and high energy electrons penetrate deeply into the material and can be used to sterilize large batches of masks within a short time period. In relation to reports that sterilization by ionizing radiation reduces filtration efficiency of polypropylene membrane filters on account of static charge loss, we have demonstrated that both gamma and electron beam irradiation can be used for sterilization, provided that the respirators are recharged afterwards. Image 1 • Gamma irradiation and electron beam irradiation have comparable effects on filtration efficiency of the respirators. • Sterilization with gamma and e-beam irradiation doesn't cause chemical changes in polypropylene membranes used in respirators. • The respirators can be reused after sterilization if charge is re-applied. [ABSTRACT FROM AUTHOR]