Dismantling the Nuclear Research Reactor Thetis

2013 ◽  
Author(s):  
P. Michiels
Keyword(s):  
High Speed ◽  
Research Reactor ◽  
Ventilation System ◽  
Reactor Core ◽  
Nuclear Research ◽  
Ground Level ◽  
Pneumatic Transport ◽  
Fissile Material ◽  
Storage Room ◽  
Water Cleaning

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation-analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage-room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written).

scholarly journals Design of a FPGA-based controller for power and period measurement in the start range of Dalat Nuclear Research Reactor

2021 ◽  
Vol 10 (3) ◽  
pp. 41-48
Author(s):  
Vo Van Tai ◽  
Nguyen Van Kien ◽  
Nguyen Nhi Dien ◽  
Trinh Dinh Hai ◽  
Le Van Diep
Keyword(s):  
Neutron Flux ◽  
High Speed ◽  
Research Reactor ◽  
Full Range ◽  
Reactor Core ◽  
Nuclear Research ◽  
Reactor Power ◽  
Processing Unit ◽  
Upward Movement ◽  
Nuclear Research Reactor

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.

A Research Reactor Source Term and Dose Results in the Event of a LOCA

2014 ◽  
Author(s):  
Charalampos Pappas ◽  
Andreas Ikonomopoulos ◽  
Athanasios Sfetsos ◽  
Spyros Andronopoulos ◽  
Melpomeni Varvayanni ◽  
...  
Keyword(s):  
Research Reactor ◽  
Source Term ◽  
Dispersion Model ◽  
Ventilation System ◽  
Atmospheric Dispersion ◽  
Reactor Core ◽  
Forecast Model ◽  
Scale Model ◽  
Continuous Reactor ◽  
Reactor Building

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

scholarly journals Confocal chromatic sensor for displacement monitoring in research reactor

2021 ◽  
Vol 253 ◽  
pp. 04021
Author(s):  
Marion Agoyan ◽  
Gary Fourneau ◽  
Guy Cheymol ◽  
Ayoub Ladaci ◽  
Hicham Maskrot ◽  
...  
Keyword(s):  
Small Volume ◽  
Research Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Displacement Monitoring ◽  
Sensor Performance ◽  
Fuel Rod ◽  
Displacement Thickness ◽  
Nuclear Research Reactor ◽  
Mitigation Techniques

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.

scholarly journals GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM

2012 ◽  
Vol 14 (2) ◽  
Author(s):  
Syarip Syarip ◽  
Widyatmaka Susyanta ◽  
Hadi Kusuma
Keyword(s):  
Fission Product ◽  
Radiation Exposure ◽  
Performance Improvement ◽  
Research Reactor ◽  
Source Term ◽  
Ventilation System ◽  
Ground Level ◽  
Safety Performance ◽  
Ground Level Concentration ◽  
Reactor Building

GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM. The safety performance of Kartini research reactor related to the gaseous releases to the environment has been evaluated. The research covers an evaluation and improvement on the ventilation system and analysis of gas releases dissipating from the reactor building. The method used is calculation of reactor source term and direct measurement of gas release from the reactor stack. The source term analysis showed that the fission product accumulated in the reactor core at the start of operation was 4.838 ´ 106 Ci, after of 5 hours operation it became 3.614 ´ 108 Ci, and after 24 hours decay, the fission product became 4.727 ´ 106 Ci. The N16 activity inside the reactor building is 4.1 ´ 10-10 μCi/cm3 and the Ar41 escaping to the atmosphere is 5.7 ´ 10-12 mCi/cm3, which is lower than limit value for radiation worker of 2 ´ 10-6 μCi/cm3. A sample case by using March 2009 data, the value of ground level concentration on variable distance x = 100 m to 5.000 m, was 9.726 ´ 10-19 rad/m3, rise up to 6.303 ´ 10-14 rad/m3 and tends to decrease to 1.598 ´ 10-15 rad/m3 at distance 5,000 m. Whiles the direct observation on the upper reactor stack show that the radiation exposure is 2.33 ´ 10-9 rad/s, exit velocity of gas from stack is 8 m/s, absolute temperature effluent of gas is 26.2 oC, and outlet diameter of stack, d = 1 m and actual stack height 31.75 m. Based on safety limit criteria from national regulation (BAPAETEN), the values of radiation exposure, ground level concentration combined with atmosphere stability and demography factor was very safe for the actual condition of Kartini reactor site. Keywords: safety performance, Kartini reactor, source term, ventilation system.

Innovation of Irradiation Equipment at the LVR-15 Nuclear Research Reactor

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Zdena Lahodová ◽  
Witolda Soukupová ◽  
Michal Koleška ◽  
Jaroslav Ernest ◽  
Jelena Zmítková
Keyword(s):  
Research Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Entire Length ◽  
Research Center ◽  
Irradiation Condition ◽  
Specimen Holder ◽  
Nuclear Research Reactor ◽  
Radiation Research

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.

Thermal-hydraulic analysis of a research reactor on personal computer with TARR code

1992 ◽  
Vol 14 (3) ◽  
pp. 1-5
Author(s):  
Ngo Huy Can ◽  
Nguyen Manh Lan ◽  
Tran Van Tran
Keyword(s):  
Personal Computer ◽  
Research Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Stationary Regime ◽  
Hydraulic Calculation ◽  
Hydraulic Analysis ◽  
Fuel Assemblies ◽  
Nuclear Research Reactor ◽  
Thermal Hydraulic Analysis

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.

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