SISTEM INFORMASI TATA KELOLA PERPUSTAKAAN BERBASIS WEB PADA DINAS KEARSIPAN DAN PERPUSTAKAAN PROVINSI GORONTALO

The research aims to facilitate the process of processing library data in the form of database storage, resulting in accurate reports, in pandemic times where all employees are required to do their duties from home so that staff in the field of services and technology, staffing, and governance have difficulty processing library data. Although it has used a computer as a means of support, the processing of data libraries is still stored using Ms.office, resulting in ineffectiveness in the work. The application of the information system has helped employees in accelerating the creation and submission of reports carried out in the Ministry of Archives and Library of Gorontalo Province using descriptive methods. The number of visitors who come on average is 51 people/day, and for the month as many as1,224 people. So the total number of visitors in 2020 is 14,688 people.Web-Based Library Governance Information System at the Ministry of Archives and Library of Gorontalo Province can provide convenience in processing Data Library and assist employees in doing their work from home (WFH)

Cross-section measurements of the (n,2n) and (n,p) reactions on 124,126,128,130,131,132Xe in the 14 MeV region and theoretical calculations of their excitation functions

The reaction cross-sections of 124Xe(n, 2n)123Xe, 126Xe(n, 2n)125Xe, 128Xe(n, 2n)127Xe, 130Xe(n, 2n)129mXe, 132Xe(n, 2n)131mXe, 130Xe(n, p)130I, 131Xe(n, p)131I, and 132Xe(n, p)132I were measured at the 13.5, 13.8, 14.1, 14.4, and 14.8 MeV neutron energies. The monoenergetic neutrons were generated through the 3H(d,n)4He reaction at the China Academy of Engineering Physics using the K-400 Neutron Generator with a solid 3H-Ti target. A high-purity germanium detector was used to measure the activities of the product. The reactions 93Nb(n, 2n)92mNb and 27Al(n, α)24Na served for neutron flux calibration. The cross sections of the (n,2n) and (n,p) reactions of the xenon isotopes were acquired within the 13–15 MeV neutron energy range. These cross-sections were then compared with the IAEA-exchange format (EXFOR) database-derived experimental data together with the evaluation results of the CENDL-3, ENDF/B-VIII.0, JENDL-4.0, RUSFOND, and JEFF-3.3 data libraries as well as the theoretical excitation function obtained using the TALYS-1.95 code. The cross-sections of the reactions (except for the 124Xe(n, 2n)123Xe and 132Xe(n, p)132I) at 13.5, 13.8, and 14.1 MeV are reported for the first time in this work. The present results are helpful to provide better cross-section constraints for these reactions in the 13–15 MeV region, thus improving the quality of the corresponding database. Meanwhile, these data can also be used for the verification of relevant nuclear reaction model parameters.

Impact of Different Nuclear Data Library on Control Rod Reactivity Worth Calculation of Small Pebble Bed Reactor

One of the main critical issues on a nuclear reactor is safety and control system. The control rod worth plays an important role in the safety and control of nuclear reactors. The control rods worth calculation is used to specify the safety margin of the reactor. The main objective of this work is to investigate impact of different nuclear data libraries on calculating the control rod reactivity worth on small pebble bed reactor. Calculation of the control rod reactivity worth in small high temperature gas cooled reactor has been conducted using the Monte Carlo N-Particle 6 (MCNP6) code coupled with a different nuclear data library. Famous evaluated nuclear data libraries such as JENDL-40u, ENDF/B-VII.1 and JEFF-3.2 continuous cross section-energy data libraries were used. The overall calculation results of integral control rod worth show that the ENDF/B-VII.1, JENDL-40u and JEFF-3.2 files give values of - 17.814%☐k/k, -18.0204 %☐k/k and -18.0267%☐k/k, respectively. Calculations using ENDF/B-VII.1 give a slightly lower value than the others, while the JENDL-4.0u file gives results that are close to JEFF-3.2 file. The different nuclear data libraries have a relatively small impact on the control rod worth of small pebble bed reactor. Accurate prediction by simulation of control rod worth is very important for the safety operation of all reactor types, especially for new reactor designs.

Calculation of neutronic characteristics in different reflector materials with a 15-MWt reactor core using VVR-KN fuel type

VVR-KN is one of the low enriched fuel types intended for a research reactor of a newCentre for Nuclear Energy Science and Technology (CNEST) of Viet Nam. As a part of design orientation for the new research reactor, the calculations of neutronic characteristics in a reactor core reflector using different materials were carried out. The investigated core configuration is a 15-MWt power loaded with VVR-KN fuel assemblies and surrounded by a reflector using beryllium, heavy water or graphite respectively. MCNP5 code together with up-to-date nuclear data libraries were used for these calculations. This paper presents the calculation results of neutron energy spectrum, neutron spatial distribution in the reflector using the above-mentioned materials. Besides, neutronic characteristics calculated for silicon doping irradiation holes in the reflector are also presented and the utilization capabilities of different reflector materials are discussed.

Simulation of the burnup in cell calculation using the wimsd-5b code considering different nuclear data libraries

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups.