Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

2015 ◽  
Vol 85 ◽  
pp. 384-390 ◽  
Author(s):  
W. Titouche Kouidri ◽  
F. Letaim ◽  
A. Boucenna ◽  
M.H. Boulhaouchet
2021 ◽  
Vol 253 ◽  
pp. 04021
Author(s):  
Marion Agoyan ◽  
Gary Fourneau ◽  
Guy Cheymol ◽  
Ayoub Ladaci ◽  
Hicham Maskrot ◽  
...  

Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision. Its operation principle is based on a wavelength encoding of the object position. Historically, the company STIL based in the south of France has first developed this class of sensors in the 90’s. Of course, this sensor can only operate in a sufficiently transparent medium in the used spectral domain. It presents the advantage of being contactless, which is a crucial advantage for some applications such as the fuel rod displacement measurement in a nuclear research reactor core and in particular for cladding-swelling measurements. The extreme environmental conditions encountered in such experiments i.e. high temperature, high pressure, high radiations flux, strong vibrations, surrounding turbulent flow can affect the performances of this optical system. We then need to implement mitigation techniques to optimize the sensor performance in this specific environment. Another constraint concerns the small volume available in the irradiation rig next to the rod to monitor, implying the challenge to conceive a miniaturized sensor able to operate under these constraints.


2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Kien-Cuong Nguyen ◽  
Vinh-Vinh Le ◽  
Ton-Nghiem Huynh ◽  
Ba-Vien Luong ◽  
Nhi-Dien Nguyen ◽  
...  

Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4  μ Sv / h is small enough for people to work normally.


2021 ◽  
Vol 10 (3) ◽  
pp. 41-48
Author(s):  
Vo Van Tai ◽  
Nguyen Van Kien ◽  
Nguyen Nhi Dien ◽  
Trinh Dinh Hai ◽  
Le Van Diep

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.


Author(s):  
Zdena Lahodová ◽  
Witolda Soukupová ◽  
Michal Koleška ◽  
Jaroslav Ernest ◽  
Jelena Zmítková

This paper describes the design and use of a new irradiation facility for the LVR-15 nuclear research reactor. The CHOUCA MT irradiation rig was produced in France according to a design of the ÚJV Group (ÚJV Řež and Research Center Řež). There are six heating sections situated along the rig, each instrumented and controlled by its own thermocouple. The rig’s insulation layers ensure a balanced temperature in irradiated specimens along its entire length. The specimen holder is 55.9 mm in diameter and 320 mm long. The CHOUCA MT rig can be repetitively irradiated in different positions within the reactor core, depending on irradiation condition requirements. The CHOUCA MT rig expands the possibilities of radiation research in the ÚJV Group.


2021 ◽  
Vol 10 (4) ◽  
pp. 08-15
Author(s):  
Cao Thanh Long ◽  
Truong Hoang Tuan ◽  
Huynh Dong Phuong ◽  
Nguyen Hoang Nhat Khang ◽  
Ho Manh Dung

A PC-based real-time interactive transient simulator of Dalat Nuclear Research Reactor (DNRR), namely DalatSim, based on the best-estimate thermal-hydraulic code RELAP5/MOD3.3 has been currently building at Center for Nuclear Technologies (CNT). This paper presents the study on developing the physics core, control module, and human-machine interface (HMI) of DalatSim. The nodalization of DNRR used for DalatSim was based on the reported numerical model in the Safety Analysis Report (SAR) in 2012. DalatSim can simulate operational procedures and several hypothetical transient accidents of DNRR. A curve of real operational power of DNRR was used to compare with calculation power results from DalatSim to verify its capability. The verification results are presented and discussed.


1992 ◽  
Vol 14 (3) ◽  
pp. 1-5
Author(s):  
Ngo Huy Can ◽  
Nguyen Manh Lan ◽  
Tran Van Tran

The code has been created for thermal-hydraulic calculation of stationary regime of nuclear research reactor, using personal computer. The main objective of the code is to compute the thermal parameters in the reactor core in order to avoid any accident. The code can be applied for many fuel assemblies available in research reactors.


Sign in / Sign up

Export Citation Format

Share Document