Volume 83, Issue 2

The volume porosity of irradiated UO 2 fuel in the form of a correlation is developed which includes two individual parts as the pore and swelling porosity in terms of the local density and fractional matrix swelling. Based on a low temperature swelling model for UO 2 fuel, the matrix swelling terms are calculated and the evolution of porosity up to burn-up levels of 120 MWd/KgU is presented. It is seen that for high burn-up PWR fuel pellet rim, the dominant contribution to total porosity comes from pores generated due to Xe depletion in the ceramic matrix while the nearly constant swelling porosity makes a small contribution to the total porosity. The porosity evolution is also validated by comparing against experimental data and a fairly good agreement was found.

The purpose of this paper is to provide a good knowledge and understanding on the main factors influencing human performance in complex systems operation. Identification and establishment of the necessary data for human factor performance analysis in a socio-technical context based on the analysis of HRA methods are performed. An evaluation of the techniques used for collecting and characterization of necessary data for HRA (observation, questionnaire, interview, task analysis, change analysis, simulator, expert judgement) is achieved also. The necessary data to perform HRA for the TRIGA reactor operation, for activities of the post-irradiation examination laboratory and activities of the radioactive waste plant treatment from RATEN ICN Pitesti were obtained. Recommendations on how to obtain data and knowledge related to development of methodical framework of HRA data collection, characterization and evaluation are also presented in this paper.

Innovative Research Reactor 50 MW (RRI-50) is a conceptual design of an Indonesia's high-power innovative research reactor, which is designed to produce a maximum neutron flux as high as 1.0 × 10 15 n/cm 2 s and thermal neutron flux of 5.0 × 10 14 n/cm 2 s. In this study, the design basis accident slow reactivity insertion is carried out. Reactivity insertion accident (RIA) caused by the inadvertent withdrawal of all fuel rod at normal rate is simulated by using PARET-ANL transient code. In the simulation accident is assumed to occur when the reactor is in operation at high power and at low power level. During transient, the fuel rods are assumed to be in the most effective positions with insertion rate of 0.049 $/s. This reactivity worth will give effect only if the fuel rod interlock system is failed to function properly and the operator action does not follow the standard operation procedure. The analysis results show that the reactor experiences scram when it reaches its protected power at high power level of 110 % (55 MW) at period of 5 s and at low power level of 3 MW at period of 120 s, and there is sufficient safety margin for the anticipated RIA caused by reactivity feedback.

CANDU reactor flexibility in using advanced fuel options allowed the continuous improvement of its performances. Both economic benefits and better operational safety are expected from thorium-based fuel alternatives. This paper addresses the use of uranium – thorium mixed oxide in 37- and 43-element CANDU fuel bundles, in the aim of obtaining better economic and safety performances. Thorium is expected to partially replace uranium in the fuel, the expected Th/U ratio being 1.5 ÷ 2.333. These bundles are meant to be loaded in every channel of the CANDU core, in a homogenous “once-through Thorium” (OTT) core. The low-enriched uranium (LEU) with 5 ÷10 wt% U-235 is a suitable fissile driver material for thorium in CANDU, allowing the chain reaction to be initiated throughout the core. Cell calculations were performed using the transport codes DRAGON and WIMS, as well as the associated nuclear data library, in order to estimate the key fuel performance parameters: maximum burnup, power peaking factors, fissile utilization, void reactivity, Doppler effect and proliferation resistance.

The aim of this study is to determine uranium ( 238 U), potassium ( 40 K) and thorium ( 232 Th) in the radioactive elements by measuring the natural rocks in the vicinity of Antalya province by using High-purity Germanium (HPGe) Detector with gamma spectrometry 69. At the end of the measurements, total absorbed dose rates of the rock samples have been found to be within the normal range of values between 3.10 nGy/h and 117.35 nGy/h. While the minimum value of Raeq has been determined to be 6.36 Bq/kg at sample S8, the maximum value has been determined as 250.15 Bq/kg at sample S4. All of the values obtained at the end of the study have been determined to be lower than the maximum Raeq value (370 Bq/kg) and the limit values for industrial products which ranges between 370 – 740 Bq/kg. ARa, ATh and AK; 232 Th, 40 K and 226 Ra of radionuclides are the activity concentrations in terms of Bq/kg. The hex value should be less than 1. The value is based on the upper limit of Raeq (370 Bq/kg). All of the results obtained are less than the limit value of 1.

One of the most important technical documents for obtaining of a reactor license (site selection, design, construction and operation license) from IINRA is public dose assessment and calculation of discharge limits into the environment. Presence of population centers and aquifers adjacent of nuclear installations is a crucial rule in public dose assessment and determination of liquid release limits into the environment. In this paper, liquid waste of TRR regards as liquid released in sewage system. The migration of radionuclides released to well through the ingestion of radionuclides via drinking water and human food-chain, inhalation of radionuclides and external irradiation from radionuclides in soil and water was modeled using the AMBER software. Then, amount of liquid released to environment with different amount of distribution coefficient, as an important effective parameter in simulation, is determined. The calculation of discharge limits shows that the liquid radioactive waste is lower than the values given into the Environmental Assessment for Tehran Research Reactor report. So the received dose doesn't exceed the dose limit of 10 μSvyr −1 for members of the public which is approved by INRA.

The recycling of the spent fuel of nuclear reactors is the important aspects in the fuel cycle technology. One of the recycling of the spent fuel is reuse of spent PWR (pressurized water reactor) fuel in CANDU reactor after re-fabricated into CANDU (CANda deuterium uranium) fuel bundles; this technology called Direct Use of Spent PWR Fuel In CANDU Reactor (DUPIC). This technology was modified however; it introduced a new design for the PWR fuel assembly. The new fuel can be transferred directly to CANDU reactors without processing, only after a certain cooling time. In this study the evaluation from thermal hydraulic point of view is done. A Westinghouse PWR fuel assembly was selected as a reference. Thermal hydraulic simulation for reference and new design fuel occurred by using RELAP5 (Reactor Excursion and Leak Analysis Program) code, then comparing the both results. The hottest rod was considered a worst condition in the reactor so the model simulated it rather than whole core. The new design achieves a good results comparing with a reference one.

Errors in suppression of xenon oscillations refer to anticipated operational occurrences. Such errors could be treated as a long-term absence of some special actions used to stabilize the power field or incorrect movements of the control and protection system absorber rod. In this study, the movement of control rod, causing maximum power tilt, has been simulated in a WWER-1000 reactor by using NJOY and MCNPX codes and also a single heated channel model for thermal hydraulic calculations. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. Analyses are based on linear heat rate of the most-powered fuel rod in the beginning of cycle and end of cycle. First, we determined the position of this most-powered fuel rod at the beginning and end of cycle separately. Then the most-powered fuel rod is segmented axially into equally spaced zones to study the effect of axial linear heat rate profile. Finally, the results have been compared with Final Safety Analysis Report of WWER-1000 reactor. It is seen that there is a great similarity between calculated and reported results.

This study presents a novel method for determining the density of liquid phase in annular regime of two phase flows by utilizing dual modality densitometry technique and artificial neural network (ANN) model of radial basis function (RBF). The detection system is comprised of one 137 Cs radioactive source, one 1 inch sodium iodide (NaI) scintillation detector for registering transmitted photons and one 1 inch NaI detector for registering scattered photons. An experimental setup was designed in order to provide the required data for testing and training the ANN. Registered counts in two detectors were utilized as the inputs of ANN and density of liquid phase was used as the output of ANN. The results show that the proposed approach could be successfully applied for predicting the density of liquid phase in annular regime of gas-liquid two phase flows. The liquid phase density was predicted with root mean square error (RMSE) of less than 0.0155 without a prior knowledge of the void fraction percentage in annular regime.

The Taban device has a circular cross section chamber, which is under construction in Amirkabir University of Technology. It has been designed to work in two modes: Tokamak and simple magnetized device. The main part of the Taban device design was the Toroidal Field coils. In this paper, a refined design of the Toroidal Field coil system for the Taban device was presented. Toroidal field coil consist of 16 circular shape coils connected in series, with 30 turns each coils, its material is cooper with 6 mm 2 area (16 AWG) that can carry 3 kA transient current. In tokamak mode, that makes the magnetic field of 0.7 T at the plasma center. The total stored energy is 130 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 2.6 mF and V max = 10 kV. In simple magnetized device mode, magnetic field is 0.1 T at high field side. The stored energy is 20 kj and the power supply used in this system is capacitor bank with capacity of C = 40 mF and V max = 1 kV.