Reactor power dependency of the neutron flux parameters fork 0-standardization

1997 ◽  
Vol 221 (1-2) ◽  
pp. 241-244 ◽  
Author(s):  
O. Díaz Rizo ◽  
M. V. Manso Guevara ◽  
E. Herrera Peraza ◽  
I. Alvarez Pellón ◽  
M. C. López Reyes
Keyword(s):  
Neutron Flux ◽  
Reactor Power ◽  
Neutron Flux Parameters

Determination of the neutron flux density at an unknown reactor power

1987 ◽  
Vol 14 (12) ◽  
pp. 677-680
Author(s):  
F. Tomášek
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Reactor Power ◽  
Neutron Flux Density

Possible Implementation of Ex-core Measurement in TEPLATOR Graphite Reflector

2021 ◽  
Author(s):  
Eva Vilimova ◽  
Tomas Peltan ◽  
Jana Jiricková
Keyword(s):  
Neutron Flux ◽  
District Heating ◽  
Flux Measurement ◽  
Heating System ◽  
Fast Response ◽  
Reactor Power ◽  
Measuring System ◽  
Monte Carlo Code ◽  
District Heating System ◽  
And Control

Abstract An ex-core neutron flux measurement is a crucial system for all common power reactors. It is necessary to monitor the neutron flux and control the chain reaction, therefore the ex-core neutron flux measurement is one of the main safety and control systems. The main advantage of this arrangement of detectors is a fast response to neutron flux change, which determines the reactor power change. Regarding to the new reactor concepts, it is important to deal with improved detection systems suitable for these reactors. Many of the modern reactor concepts are based on a graphite moderator or reflector, which is also the case of the TEPLATOR. The TEPLATOR is a solution of a district heating system based on heavy water as a moderator and graphite as a reflector. The TEPLATOR is designed to use irradiated fuel from the commercial PWR or BWR reactors, which has low to intermediate burnup. This article is focused on the verification of the possible use of the special neutron measuring system placed in the graphite reflector. The Monte Carlo code Serpent was used for the calculations performed in this article.

Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

2010 ◽  
Author(s):  
Aimin Zhang ◽  
Yalun Kang
Keyword(s):  
Fuel Element ◽  
Neutron Flux ◽  
Research Reactor ◽  
High Heat ◽  
Reactor Power ◽  
High Heat Flux ◽  
Water Tank ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Plate Type

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Neutron flux parameters at irradiation positions in the new research reactor FRM-II

2006 ◽  
Vol 564 (2) ◽  
pp. 641-644 ◽  
Author(s):  
Xilei Lin ◽  
Richard Henkelmann ◽  
Andreas Türler ◽  
Heiko Gerstenberg ◽  
Frans De Corte
Keyword(s):  
Neutron Flux ◽  
Research Reactor ◽  
Neutron Flux Parameters ◽  
New Research

scholarly journals On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Luigi Lepore ◽  
Romolo Remetti ◽  
Mauro Cappelli
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Signal To Noise Ratio ◽  
Reactor Power ◽  
Fast Reactors ◽  
Flux Monitoring

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia

2012 ◽  
Vol 70 (10) ◽  
pp. 2488-2493 ◽  
Author(s):  
A.R. Yavar ◽  
H. Khalafi ◽  
Y. Kasesaz ◽  
S. Sarmani ◽  
R. Yahaya ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Neutron Flux Parameters ◽  
Mcnp Simulation

scholarly journals Advanced Test Reactor: A National Scientific User Facility

2008 ◽  
Author(s):  
Clifford J. Stanley ◽  
Frances M. Marshall
Keyword(s):  
Neutron Flux ◽  
National Laboratory ◽  
Reactor Core ◽  
Nuclear Research ◽  
Reactor Power ◽  
Department Of Energy ◽  
Life Extension ◽  
Test Environment ◽  
User Facility ◽  
Test Reactor

This presentation and associated paper provides an overview of the research and irradiation capabilities of the Advanced Test Reactor (ATR) located at the U.S. Department of Energy Idaho National Laboratory (INL). The ATR which has been designated by DOE as a National Scientific User Facility (NSUF) is operated by Battelle Energy Alliance, LLC. This paper will describe the ATR and discuss the research opportunities for university (faculty and students) and industry researchers to use this unique facility for nuclear fuels and materials experiments in support of advanced reactor development and life extension issues for currently operating nuclear reactors. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration (Fig. 1) of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and chemistry can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ∼1.0 x1015 n/cm2-sec with a maximum fast flux of ∼5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

Research Reactor VR-1 and Determination of Neutron Spatial Distribution in Its Radial Experimental Channel

2009 ◽  
Author(s):  
Marija Mileticˇ
Keyword(s):  
Spatial Distribution ◽  
Neutron Flux ◽  
Flux Density ◽  
Research Reactor ◽  
Reactor Power ◽  
Activation Detector ◽  
Experimental Channel ◽  
Neutron Flux Density ◽  
Radial Channel

The research reactor VR-1 is operated by Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University (CTU) in Prague. It is a pool-type, light-water reactor, with low enriched uranium. Maximum thermal power is 1kW (equal to 1·108 impulses/second when compared with reactors with higher power). Research on VR-1 reactor is mainly used for the education of university students, preparation and testing of new educational methodologies, investigation of reactor lattice parameters, reactor dynamics study, research in the control equipment field, neutron detector calibration, etc. One of the applications performed by students is the determination of the absolute value of the neutron flux density (also known as Neutron Spatial Distribution) in the radial experimental channel in reactor VR-1. The method used for this measurement is Neutron Activation Analysis. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector which is irradiated in the experimental channel. The activity of the produced radioactive products (radioisotopes) is then measured by means of appropriate counter system (in our case, High Purity Germanium detector). For this measurement totally 34 gold foils were irradiated at different reactor power levels and various positions in radial channel in aim to determine the neutron spatial distribution in radial channel. Interesting results about symmetry, value and dependence on reactor power level of neutron flux density were obtained.

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