Dose Rates From the Accidental Withdrawal of a Fuel Element From an Open Pool-Type Reactor

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Amr Abdelhady
Keyword(s):  
Fuel Element ◽  
Dose Rate ◽  
Human Error ◽  
Normal Operation ◽  
Control Room ◽  
Radiation Dose Rate ◽  
Dose Rates ◽  
Type Reactor ◽  
Nuclear Fuel Element ◽  
Pool Type

This paper studies the radiological consequences resulting from withdrawal of nuclear fuel element (FE) from a core of open pool type reactor during normal operation. The FE withdrawal accident may be occurred due to human error during routine transport of spent FE process or from failure of FE clamp during reactor normal operation. For both cases, the FE will move vertically upward toward pool water surface. In case of accidental failure of fuel clamp, a negative reactivity insertion in the core after FE withdrawal and the reactor will be shutdown. MCNP5 code was used in this study to calculate the radiation dose rate levels in the reactor hall and inside the control room during FE withdrawal. The results show that the operator in control room will receive dose rate lower than permissible dose rate limit when FE reaches at depths more than to 211 cm for vertical FE and at depths more than to 246 cm for horizontal FE.

scholarly journals EVALUATION OF RADIATION DOSE RATE OF RSG-GAS REACTOR

2018 ◽  
Vol 24 (3) ◽  
Author(s):  
Amir Hamzah ◽  
Hery Adrial ◽  
Subiharto Subiharto
Keyword(s):  
Radiation Dose ◽  
Radiation Exposure ◽  
Dose Rate ◽  
Nuclear Reactor ◽  
Area Measurement ◽  
Radiation Dose Rate ◽  
Safety Level ◽  
Modeling And Analysis ◽  
Dose Rates ◽  
Limit Value

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.

Reliability in Microjoining Processes for Instrumented Nuclear Fuel Element Fabrication

2020 ◽  
Vol 1157 ◽  
pp. 31-37
Author(s):  
Călin Truţă ◽  
Adrian Amzoi ◽  
Dumitru Barbos
Keyword(s):  
Fuel Element ◽  
Nuclear Fuel ◽  
Structural Reliability ◽  
Added Value ◽  
Impedance Match ◽  
Normal Operation ◽  
Local Loss ◽  
Steel Coil ◽  
Nuclear Fuel Element ◽  
Prolonged Contact

The paper presents the assembling flux of thermocouple-instrumented nuclear fuel element for research reactor, from the point of view of the welding / brazing engineer, considering nuclear quality and safety requirements: fuel element structural reliability (no radioactive leaks through joints) and temperature signal reliability (thermocouple sheath integrity), this signal being an essential parameter for reactor normal operation and emergency shut-down. The paper is a real case study for an experimental instrumented element recently developed at INR-Pitesti describing technology choices as balance between fabrication complexity and risk of failure in joining processes, especially in later stages when added value increases. All joints (welded or brazed) fall into microjoining category, and it is shown how some special provisions may influence reliability. Focus is put on brazing thin-walled Inconel sheathed thermocouples, where erosion and local loss of ductility are known issues, leading to sheath rupture. Choosing as filler the less aggressive BNi-9 helped too little. A simple but efficient technique has been developed to address this matter adequate to narrow spaces inside a nuclear fuel element, where no room is available for solutions described in literature e.g. distal preplacing of filler. The solution prevents sheath from having prolonged contact with large volume of molten filler by using locally a miniature barrier (thin stainless-steel coil or sleeve) which only allows capillary wetting, being also a perfect real-time visual indicator of brazing progress and completion. As proved in the present paper, this method along with using filler formulation with lower Carbon content (without organic binder) enhances significantly, 8 times at least, resistance to bending fatigue. A particular vacuum brazing chamber design is employed: narrow quartz tube with external induction coil and top fitting letting outside the long thermocouples attached, reducing much the chamber volume and degassing. Careful impedance match is therefore required to overcome induction power loss due to the larger coil-to-workpiece gap. Additional joining problems are discussed e.g. inherent differential expansion of long parts during induction heating which afterwards may put tension upon braze during solidification and determine delayed cracking, this being avoided through wise order of operations. Another concern is the final precision weld between instrumentation segment having attached the hard-to-handle long thermocouples bunch and nuclear segment with the heavy Uranium pellets. The result of this research is successful assembling of first Romanian prototype with joints exhibiting He leak rate bellow 1E-09 std.cc/sec and overall reliability proved during reactor irradiation testing.

Development of New Simulation Software System to Evaluate Radiation Dose Rates in a Wide Radioactively Contaminated Area

2012 ◽  
Author(s):  
Tomoharu Hashimoto ◽  
Masahiro Kondo ◽  
Ryuichi Tayama ◽  
Hideho Gamo
Keyword(s):  
Radiation Dose ◽  
Dose Rate ◽  
Gamma Ray ◽  
Simulation Software ◽  
Radiation Dose Rate ◽  
Dose Rates ◽  
Reduce Radiation Exposure ◽  
Radiation Sources ◽  
Reduction Effect ◽  
Contaminated Area

The Japanese government plans to conduct decontamination tasks in radioactively contaminated areas. For such a situation, we developed a system that evaluates radiation dose rates in a wide radioactively contaminated area by utilizing our radiation dose evaluation technology. This system can not only generate present maps of radiation dose rate in the air based on the dose rate measured at the surface of the contaminated areas, but can also quickly calculate the reduction effect of dose rate due to decontamination tasks by entering decontamination factors. The system can then formulate decontamination plans and make it possible to plan measures to reduce radiation exposure for workers and local residents. Radioactive nuclides that contribute to gamma-ray dose rate are mainly Cs-134 and Cs-137 in soil, on trees, buildings, and elsewhere. Shapes of such radiation sources are assumed to be 10m square or 100m square. If it is unsuitable that the radiation sources assume to squares, the radiation sources can assume to point. The relation between distance from the surface or point source and the radiation dose rate is calculated using MCNP5 code (A General Monte Carlo N-Particle Transport Code - Version 5), and approximated using four-parameter empirical formula proposed by Harima et al. In addition, the system can consider shielding such as soil, concrete, and iron. When setting such shielding, the skyshine dose rate is taken into account in dose rate calculation.

scholarly journals Chernobyl-level radiation exposure damages bumblebee reproduction: a laboratory experiment

2020 ◽  
Vol 287 (1937) ◽  
pp. 20201638
Author(s):  
Katherine E. Raines ◽  
Penelope R. Whitehorn ◽  
David Copplestone ◽  
Matthew C. Tinsley
Keyword(s):  
Radiation Dose ◽  
Radiation Exposure ◽  
Dose Rate ◽  
Colony Growth ◽  
High Dose ◽  
Radiological Protection ◽  
Negative Consequences ◽  
Exclusion Zone ◽  
Radiation Dose Rate ◽  
Dose Rates

The consequences for wildlife of living in radiologically contaminated environments are uncertain. Previous laboratory studies suggest insects are relatively radiation-resistant; however, some field studies from the Chernobyl Exclusion Zone report severe adverse effects at substantially lower radiation dose rates than expected. Here, we present the first laboratory investigation to study how environmentally relevant radiation exposure affects bumblebee life history, assessing the shape of the relationship between radiation exposure and fitness loss. Dose rates comparable to the Chernobyl Exclusion Zone (50–400 µGy h −1 ) impaired bumblebee reproduction and delayed colony growth but did not affect colony weight or longevity. Our best-fitting model for the effect of radiation dose rate on colony queen production had a strongly nonlinear concave relationship: exposure to only 100 µGy h −1 impaired reproduction by 30–45%, while further dose rate increases caused more modest additional reproductive impairment. Our data indicate that the practice of estimating effects of environmentally relevant low-dose rate exposure by extrapolating from high-dose rates may have considerably underestimated the effects of radiation. If our data can be generalized, they suggest insects suffer significant negative consequences at dose rates previously thought safe; we therefore advocate relevant revisions to the international framework for radiological protection of the environment.

scholarly journals Environmental dose rate assessment of ITER using the Monte Carlo method

2014 ◽  
Vol 29 (1) ◽  
pp. 34-39
Author(s):  
Alireza Karimian ◽  
Amir Beheshti ◽  
Mohammadreza Abdi ◽  
Iraj Jabbari
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Dose Rate ◽  
Equivalent Dose ◽  
Photon Radiation ◽  
Normal Operation ◽  
Whole Body ◽  
Dose Rates ◽  
The Monte Carlo Method ◽  
Radiosensitive Organs

Exposure to radiation is one of the main sources of risk to staff employed in reactor facilities. The staff of a tokamak is exposed to a wide range of neutrons and photons around the tokamak hall. The International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion engineering project and the most advanced experimental tokamak in the world. From the radiobiological point of view, ITER dose rates assessment is particularly important. The aim of this study is the assessment of the amount of radiation in ITER during its normal operation in a radial direction from the plasma chamber to the tokamak hall. To achieve this goal, the ITER system and its components were simulated by the Monte Carlo method using the MCNPX 2.6.0 code. Furthermore, the equivalent dose rates of some radiosensitive organs of the human body were calculated by using the medical internal radiation dose phantom. Our study is based on the deuterium-tritium plasma burning by 14.1 MeV neutron production and also photon radiation due to neutron activation. As our results show, the total equivalent dose rate on the outside of the bioshield wall of the tokamak hall is about 1 mSv per year, which is less than the annual occupational dose rate limit during the normal operation of ITER. Also, equivalent dose rates of radiosensitive organs have shown that the maximum dose rate belongs to the kidney. The data may help calculate how long the staff can stay in such an environment, before the equivalent dose rates reach the whole-body dose limits.

IMPORTANCE OF SYNERGISM FOR DEGRADATION OF ELASTOMERS IN COMBINED RADIATION PLUS TEMPERATURE ENVIRONMENTS

2020 ◽  
Vol 93 (1) ◽  
pp. 121-141 ◽  
Author(s):  
Kenneth T. Gillen
Keyword(s):  
Dose Rate ◽  
Rate Constant ◽  
Synergistic Effects ◽  
Chemical Kinetic ◽  
High Dose ◽  
Thermal Limit ◽  
Radiation Dose Rate ◽  
New Approach ◽  
Dose Rates ◽  
Elastomeric Materials

ABSTRACT Understanding the importance of synergistic effects that occur in combined radiation (dose rate R) plus thermal aging (temperature T) environments has been of interest for many years. Although suggested approaches for achieving this objective are contained in two recent international publications, the validity of these approaches is questioned. A new approach is described based on chemical kinetic principles, and applied to elongation data for two elastomeric materials: a chloroprene and a chlorosulfonated polyethylene. At low temperatures and high dose rates, the chemistry from radiation-initiation totally dominates the degradation, leading to the rate constant kR in the radiation limit. This rate constant is shown to be independent of temperature as long as the dose rate at that temperature is high enough to reach radiation-limit conditions. For combined environment experiments at R + T, this result, added to the rate constant kT appropriate to the thermal limit (obtained from thermal-only exposures), leads to the combined environment rate constant expected in the absence of synergism. Comparing this result to the experimental combined environment rate constant kR+T leads to quantitative estimates of synergism across (R-T) space. Important synergistic effects are found for both materials.

A National Perspective: Establishing and Implementing a Characterization Program in the U.S. for Remote-Handled Transuranic Waste

2007 ◽  
Author(s):  
Roger Nelson ◽  
Alton D. Harris
Keyword(s):  
Dose Rate ◽  
Waste Disposal ◽  
Environmental Protection Agency ◽  
Department Of Energy ◽  
Surface Dose ◽  
Radiation Dose Rate ◽  
Characterization Methods ◽  
Dose Rates ◽  
Transuranic Waste ◽  
The U.S

The U.S. Department of Energy (DOE) is responsible for waste management from nuclear weapons production and operates the Waste Isolation Pilot Plant (WIPP) for permanent disposal of defense-generated transuranic waste (TRU), as authorized by Congress in 1979. Radioactive waste in the U.S. has historically been managed in one of two ways depending on its penetrating radiation dose rate. Waste with surface dose rates above 200 millirem/hour (0.002 sievert/hour) and waste that has been managed remotely (remote-handled). In 1992, Congress passed the WIPP Land Withdrawal Act, which created the regulatory framework under which DOE was to operate the facility, and authorized disposal of waste up to 1,000 rems/hour (10 Sievert/hour). Subsequently, DOE submitted applications to the Environmental Protection Agency (EPA), at the Federal level, for certification to operate WIPP, and to the New Mexico Environment Department (NMED), at the State level, for a hazardous waste permit. Both applications described the characterization methods that DOE proposed to use to ensure only compliant waste was shipped to WIPP. No distinction was employed in these methods concerning the surface dose rate from the waste. During the applications review, both regulatory agencies came to the conclusion in their approval that DOE had not demonstrated that remote-handled transuranic (RH-TRU) waste could be adequately characterized. Therefore, WIPP was only granted approval to begin waste disposal operations of waste with surface dose rates less than 200 millirem/hour (0.002 sievert/hour) — or contact-handled transuranic (CH-TRU) waste. Emplacement of CH-TRU waste in WIPP began March 26, 1999. However, WIPP was designed for disposal of both CH- and RH-TRU waste, with the RH-TRU waste in canisters emplaced in the walls of the underground disposal rooms and CH-TRU waste in containers in the associated open drifts. Therefore, as disposal rooms filled with CH-TRU waste, the space along the walls for RH-TRU waste disposal was lost. This made removal of the regulatory prohibition on RH-TRU waste a very high priority, and DOE immediately began an iterative process to change the two regulatory bases for RH-TRU waste disposal. These changes focused on how DOE could rely on CH-TRU characterization methods for adequate characterization of RH-TRU waste. On January 23, 2007, the first shipment of RH-TRU waste was finally received at WIPP. The revised EPA certification and NMED permit now both consider all waste characterization methods to be equally effective when applied to either CH- or RH-TRU waste, as DOE maintained in the original applications over 10 years ago.

scholarly journals Application of MCP-N (Lif: Mg, Cu, P) TL detectors in monitoring environmental radiation

2004 ◽  
Vol 19 (1) ◽  
pp. 20-25 ◽  
Author(s):  
Pavel Olko ◽  
Maciej Budzanowski ◽  
Pavel Bilski ◽  
Snezana Milosevic ◽  
Barbara Obryk ◽  
...  
Keyword(s):  
Dose Rate ◽  
Field Measurements ◽  
Depleted Uranium ◽  
Short Term ◽  
Environmental Radiation ◽  
Natural Radiation ◽  
Radiation Dose Rate ◽  
Radiation Background ◽  
Thermoluminescent Dosimetry ◽  
Dose Rates

Thermoluminescent MCP-N detectors based on LiF:Mg,Cu,P are by about 2 orders of magnitude more sensitive than TLD-100 detectors based on conventional LiF:Mg,Ti, which makes it possible to use them in short-term monitoring of ionizing radiation in the environment (e. g., over a two-week period, rather than over 3-12 months). We describe the properties of MCP-N detectors and methods of their application in environmental monitoring. The system was tested in short and long-term exposure periods at 100 sites around Krakow region. MCP-N detectors were then applied to measure variation of radiation dose rate at four selected villages in Serbia, where depleted uranium ammunition was deployed in 1999. Together with short-term thermoluminescent dosimetry, in situ measurements using proportional counters were per formed in order to assess the range of variation of natural radiation background in these villages. The mean terrestrial kerma dose rate in these villages was found to vary between 85 and 116 nGyh?1 and the average ambient dose equivalent rate H*(10) determined by thermoluminescent detectors and by proportional counter measurements was 160 nSvh?1. These values of natural radiation back ground dose rates can be applied as reference levels for field measurements around other sites where depleted uranium ammunition was deployed.

Sign in / Sign up

Export Citation Format

Share Document