Monte Carlo analysis of KRITZ-2 critical benchmarks on the reactivity temperature coefficient using ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries

Ouahdani, S. El; Boukhal, H.; Erradi, L.; Chakir, E.; Bardouni, T. El; Hajjaji, O. El; Boulaich, Y.; Benaalilou, K.; Kaddour, M.

doi:10.1016/j.anucene.2015.07.010

Cited by 8 publications

(3 citation statements)

References 3 publications

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“…The results showed a very good agreement between the considered libraries in the two configurations of KRITZ-2:13, with big deviations between them and the benchmark values. The absolute values of the RTC obtained in our study for both evaluations JENDL-4.0 and ENDF/B-VII.I are lower than of other study [14] whose results were (-1.49 and -0.88) pcm/°C respectively for JENDL-4.0 and ENDF/B-VII.I. In that study the data was treated using the assistant code Makxsf (interpolation between two temperatures) along with the NJOY code.…”

Section: Critical Calculationscontrasting

confidence: 81%

“…NJOY system [12] has been used to convert the ENDF6 data of the libraries into ACE format necessary for MCNP code, as well as to generate the covariances, by means of ERRORJ module. Sensitivities 𝐺 and covariances 𝑀 will be combined according to the first equation in order to calculate the nuclear uncertainty on Keff: For a temperature range [T1, T2], with neutron factors [Keff1, Keff2], the RTC could be obtained by the following formula [14]:…”

Section: Methodsmentioning

confidence: 99%

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Cross sections Data Adjustment For KRITZ-2:13

Ahmed

Boukhal

Chakir

et al. 2021

KJS publishes peer-review articles in Mathematics, Computer Sci

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Over the past years, the cross-sections reaction data has been re-evaluated several times, in order to approximate the nuclear model measurements with the predictions with great reliability. In our work, uncertainty analysis caused by the data on the neutron factor (Keff) and the reactivity temperature coefficient (RTC), in addition to nuclear data adjustment related to the nuclear reactor physics have been done for KRITZ-2:13 reactor, with ENDF/B - VII.1, ENDF/B - VIII.0 and JENDL - 4.0 evaluations by the nuclear code MCNP6.1. Our analysis detects that the greatest uncertainty on Keff and RTC in the studied libraries comes from the capture and fission reaction contributions respectively, for U-238 and U-235. The previous reactions and their covariances were adjusted using the generalized least squares method (GLLSM), in order to contribute to improve the data needed for neutron simulation of experiments and to ensure the installations safety, where Keff and RTC represent neutron parameters reflecting the modification effects in the data.

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Section: Critical Calculationscontrasting

confidence: 81%

Section: Methodsmentioning

confidence: 99%

Cross sections Data Adjustment For KRITZ-2:13

Ahmed

Boukhal

Chakir

et al. 2021

KJS publishes peer-review articles in Mathematics, Computer Sci

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“…Ordinary concrete contains more oxygen and silicon while heavy concrete contains more iron and barium. Compared to oxygen and silicon, the iron and barium elements have greater neutron cross-section [11][12][13] and higher density, thus their ability as a radiation shielding is much better. But heavy concrete is more expensive, so it is necessary to determine the optimal combination of shielding substances and dimensions.…”

Section: Theorymentioning

confidence: 99%

Evaluation of Radiation Dose Rate of RSG-Gas Reactor

Hamzah

Adrial

Subiharto

2018

urania

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EVALUATION OF RADIATION DOSE RATE OF RSG-GAS REACTOR. The RSG-GAS reactor has been operated for 30 years. Since the nuclear reactor has been operated for a long time, aging process on its components may occur. One important parameter for maintaining the safety level of the RSG-GAS reactor is to maintain radiation exposure as low as possible, especially in the working area. The evaluation results should be able to demonstrate that the radiation exposure of the RSG-GAS is still safe for workers, communities and the surrounding environments. The purpose of this study is to evaluate radiation exposure in the working area to ensure that the operation of RSG-GAS is still safe for the next 10 years. The scope of this work is confirming the calculation results with the measured radiation dose in the RSG-GAS reactor working area. Measurement of radiation exposure is done by using the installed equipments at some points in the RSG-GAS working area and a portable radiation exposure measurement equipment. The calculations include performance of a modeling and analysis of dose rate distribution based on the composition and geometry data of RSG-GAS by using MCNP. The analysis results show that the maximum dose rate at Level 0 m working area of RSG-GAS reactor is 3.0 mSv/h with a deviation of 6%, which is relatively close to the measurement value. The evaluation results show that the dose rate in RSG-GAS working area is below the limit value established by the Nuclear Energy Regulatory Agency of Indonesia (BAPETEN) of 10 mSv/h (for the average effective dose of 20 mSv/year). Therefore, it is concluded that the dose rate in RSG-GAS working area is safe for personnel..Kata kunci: dose rates, RSG-GAS, radiation safety, MCNP.

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