Canada Deuterium Uranium Reactor Research Articles

Petri Net Modeling for Probabilistic Safety Assessment and its Application in the Air Lock System of a CANDU Nuclear Power Plant

Abstract Historically, fault trees are extensively used in Probabilistic Safety Assessment (PSA) to model and evaluate the probability and consequence of failures of complex engineering systems, such as nuclear power plants. Scenarios of hypothetical accidents resulting in severe core damage can be developed. Petri net is another modeling technique that offers many advantages when comparing with fault trees, such as is ability to represent the time sequence of the events along with their duration. In this research, Petri net theory is extended to model system failures. The transformations required to model logic gates by Petri nets are explored and examples provided. Methods for qualitative analysis for Petri nets are presented. The application of Petri net in the airlock system of a Canada Deuterium Uranium (CANDU) reactor is explored. It is further demonstrated that Petri net can be used in both coherent and non-coherent systems.

A containment analysis for SBLOCA in the refurbished Wolsong-1 Nuclear Power Plant

A small break leading to a loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analysed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The Industry Standard Toolset (IST) codes developed by CANDU (Canadian Deuterium Uranium reactor) Owners Group and updated models including design change parameters are applied newly to the event analyses. The GOTHIC code has been used for the containment thermal-hydraulic analysis of Wolsong-1. Also, the SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. The IMOD-2 was incorporated into SMART-IST as a module dealing with chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parametric studies for the IMOD-2 model are performed to decide the analysis value set. The iodine release amount increases as the paint thickness increases. But, the iodine release amount increases as the water pH (dousing and primary heat transport (PHT)) decreases. The developed containment analysis methodology and the results of SBLOCA without Emergency Coolant Injection (ECI) are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits.

Neutron-Deuteron Elastic Scattering Measurements

Neutron-deuteron elastic scattering is important for critical systems involving heavy water, including the CANDU (Canada Deuterium Uranium) reactor. A change of the elastic scattering angular distributions included in the ENDF/B-VI.5 evaluated data file for deuterium (and retained in ENDF/B-VII) has led to significant changes in Calculation/Experiment (C/E) for LANL critical benchmark experiments containing heavy water and a small, but noticeable, change in C/E for coolant void reactivity experiments relevant for CANDU reactors. Follow-up involved first principles nuclear model calculations to obtain revised angular distributions from Faddeev 3-body theory, state of the art nucleon-nucleon forces, three-body and magnetic moment interactions. The calculated differences which are most pronounced occur in the domain just above the onset of anisotropy at 50-100 keV. Here, earlier measurements show contradictory results about the magnitude of this anisotropy. At IRMM measurements are carried out in the energy range from 100 keV to 2 MeV to test the angular distributions of neutron-deuteron elastic scattering at angles where predictions show the largest differences (backward angles). An enriched sample (99.999% purity of deuterium ), consisting of a molded polyethylene disc 70 mm in diameter and 3 mm thick, was prepared and delivered through AECL. Measurements are performed at the GELINA facility using a new setup based on detection of scattered neutrons using two HPGe detectors. In front of each detector a B4C converter is used. Meanwhile we started another project to detect the recoiling deuterium using a state of the art TPC (time projection chamber) filled with deuterated P10 gas.

DIAMETRAL CREEP PREDICTION OF THE PRESSURE TUBES IN CANDU REACTORS USING A BUNDLE POSITION-WISE LINEAR MODEL

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

Predicting Diametral Creep of the Pressure Tubes in CANDU Reactors Using Fuzzy Neural Networks

Pressure tube (PT) creep is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS) in Canada deuterium uranium reactors. PT diametral creep affects the thermal hydraulic characteristics of coolant channels and the critical heat flux (CHF). CHF is a key parameter in determining the critical channel power, which is used in the trip setpoint calculations of regional overpower protection systems. This paper aims to predict PT diametral creep using the measured signals of the HTS by applying fuzzy neural networks (FNNs) according to operating conditions. The FNN model was optimized in terms of its fuzzy rules and parameters by a genetic algorithm combined with the least-squares method. Informative data that demonstrate the system's characteristic behavior were selected to train the FNN model using the subtractive clustering method. The proposed FNN model for predicting PT diametral creep was verified using the operating data of the Wolsong Unit 1 nuclear power plant in Korea. It was known that the FNN could predict the PT diametral creep accurately. Statistical and analytical uncertainty analysis methods were applied to the models and their uncertainties were evaluated using 60 sampled training and optimization data sets, as well as two fixed test data sets.

A model approach for tritium dynamics in wild mammals

Tritium (3 H) transfer into environment must be modelled differently than the transfer of other radionuclides released from nuclear reactors because hydrogen represents the building blocks of life. A solid understanding of 3 H behaviour is essential because 3 H may be released in large quantities from CANDU (CANada Deuterium Uranium) reactors and from future thermonuclear reactors. Recently, the authors published a complex dynamic metabolic model for 3 H and 14 C transfer in farm and wild animals, but the model applications for wild biota were restricted to too few examples and mostly for 14 C transfer. In this study, the model is applied to few selected wild animals for 3 H uptake. Despite the lack of any experimental data for wild animals, the results presented in this study are less uncertain than for many other radionuclides and can provide a useful estimation for biota radioprotection.

An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor

Feasibility studies for recycling the recovered uranium from electro-refining process of pyroprocessing into a Canada Deuterium Uranium (CANDU) reactor have been carried out with a source term analysis code ORIGEN-S, a reactor lattice analysis code WIMS-AECL, and a Monte Carlo analysis code MCNPX. The uranium metal can be recovered in a solid cathode during an electro-refining process and has a form of a dendrite phase with about 99.99% expecting recovery purity. Considering some impurities of transuranic (TRU) elements and fission products in the recovered uranium, sensitivity calculations were also performed for the compositions of impurities. For a typical spent PWR fuel of 3.0 wt.% of uranium enrichment, 30 GWD/tU burnup and 10 years cooling, the recovered uranium exhibited an extended burnup up to 14 GWD/tU. And among the several safety parameters, the void reactivity at the equilibrium state was estimated 15 mk. Additionally, a simple sphere model was constructed to analyze surface dose rates with the Monte Carlo calculations. It was found that the recovered uranium from the spent PWR fuel by electro-refining process has a significant radioactivity depending on the impurities such as fission products.

Neutronic performance of CANDU reactor fuelling with ThC2/233UC2

233U isotope is used as a booster fissile fuel material in the form of mixed ThC2/233UC2 fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup. Three different fuel compositions have been used: (1) 97% ThC2+3% 233UC2, (2) 98% ThC2+2% 233UC2 and (3) 99% ThC2+1% 233UC2. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full-power operation for a period of 20 years. The criticality starts by k∞=1.541, 1.355 and 0.995 for modes of (1), (2) and (3) fuel compositions, respectively. A sharp decrease in the criticality has been observed in the first 2 years as a consequence of rapid 233U burnout fuelling with (1) and (2) modes. The criticality becomes quasi-constant after the second year and remains above k∞∼1.06 for 20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner. Very high burnup could be achieved with the same fuel materials (up to 500 000 MWday t−1), provided that the fuel rod claddings would be replaced periodically (after every 500 00 or 100 000 MWday t−1). The reactor criticality will be sufficient for fuelling with (1) and (2) modes until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically. Copyright © 2010 John Wiley & Sons, Ltd.

High Temperature Aqueous Corrosion and Deuterium Uptake of Coupons Prepared from the Front and Back Ends of Zr-2.5Nb Pressure Tubes

Abstract Pressure tubes for CANada Deuterium Uranium (CANDU) reactors are extruded from billets of Zr-2.5Nb, at a temperature of ∼815°C, and then cold drawn to give a final length of ∼6 m. The manufacturing process often results in a variation of properties along the length of a tube including grain structure, texture, dislocation density, and phase distribution. This variation affects the mechanical and deformation properties as well as the aqueous oxidation and deuterium uptake behavior along the installed pressure tube. The orientation of the installed pressure tube in the reactor, with its axial variation of properties, is an important factor in the effective optimization of its service life. This work reports on the differences in aqueous oxidation and deuterium uptake between the extruded front- and back-end sections of a number of pressure tubes. The corrosion tests were conducted in heavy water in static autoclaves at Chalk River Laboratories and in a heavy water re-circulating loop in the Halden Boiling Water Reactor. The test conditions, such as water chemistry and temperature, were similar to those in the primary heat transport system of a CANDU reactor. The results indicate that under some exposure conditions, the deuterium uptake may be up to 40 % lower for back-end coupons compared to front-end coupons. Several microstructural factors including texture, grain size, and concentrations of alloying elements may cause the observed differences in deuterium uptake. The results will be discussed within the current mechanistic understandings of Zr-2.5Nb corrosion and deuterium ingress.

DOE begins rationing helium-3

As the extent of the shortage becomes clear, an interagency task force is giving scientific users priority, but some say the material is not available at any price.

Thermal-Hydraulic Aspects of Progression to Severe Accidents in CANDU Reactors

The progression of events that develop into an accident with severe fuel or core damage in the Canada deuterium uranium (CANDU) reactor is discussed. Such events involve a number of broadly common stages in which the thermal-hydraulic behavior of the reactor fuel, fuel channels, heat transport system, and a number of key process systems governs both the rate at which severely degraded cooling conditions develop and the extent of resultant damage to the reactor core. The quantification of core damage states requires the modeling of the physical phenomena that are active in these accidents, which is a focus of this paper. As discussed in this paper, unique passive features of the CANDU reactor design have a beneficial effect in that they delay the progression of severe accidents, thereby providing ample opportunity for operator actions to stabilize the plant and mitigate the consequences. It is shown that large CANDU reactors are inherently tolerant of a prolonged loss of engineered heat sinks at decay power levels. This is because two large volumes of water (the moderator and shield water) surround the reactor core and act as in situ passive heat sinks in severe accidents. This has significant impacts on severe accident management. The pressure tube reactor design precludes melting of the core at high system pressures; that is, high-pressure melt ejection is physically impossible. In the event that severe undercooling of fuel occurs at high system pressure, a pressure tube will fail well before any significant molten fuel material can accumulate.

Towards a more comprehensive microstructural analysis of Zr–2.5Nb pressure tubing using image analysis and electron backscattered diffraction (EBSD)

Zr–2.5Nb pressure tubes used in CANDU (CANada Deuterium Uranium) reactors have a very complex microstructure, with two major crystallographic phases, α and β. These phases include a fair amount of deformation from the extrusion process and the cold working (∼25%) performed at the end of the manufacturing process. This microstructure (texture, grain aspect ratio, etc.) changes along the tube’s length and differs from tube to tube. In order to better understand the deformation mechanisms, these microstructural differences must be statistically characterized. Scanning electron microscopy combined with direct image analysis or with electron backscattered diffraction (EBSD) are good techniques for carrying out such a measurement. However it is not possible, using specimen preparation methods specific for each of these techniques, to reveal all of the grain and phase boundaries. We have thus developed post-treatment algorithms to be able to partially analyze the revealed Zr–2.5Nb microstructure. The first algorithm was used for image analysis treatments of micrographs taken at 5 kV on the radial–tangential plane of etched samples using a reactive ion etch (RIE, CF 4 + O 2). The second was developed for EBSD grain mapping and can be used to characterize α-Zr grain shape and orientation. The two techniques are complementary: EBSD gives information about the micro-texture and the relationship between the microstructure and micro-texture while image analyses of SEM micrographs reveal the direction and distribution of the α-Zr lamellae more easily and over a greater sample area than EBSD. However, the SEM micrographs that were used did not reveal any grain boundary (only phase boundary). An analysis of EBSD grain maps reveals that the average α-Zr grain size, mainly in the elongated direction (tangential), is smaller than what is normally obtained from an image analysis of SEM micrographs. The grain size distribution of type I α-Zr grains (deformed original (prior) α-Zr) and type II (stress-induced β-Zr → α-Zr phase transformation) is also shown to be different for sizes greater than 0.4 μm 2.

The ASTEC Integral Code for Severe Accident Simulation

For several years the French Institut de Radioprotection et de Sûreté Nucléire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH have been jointly developing a system of calculation codes—the integral Accident Source Term Evaluation Code (ASTEC)—to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment, i.e., the source term. ASTEC has progressively reached a larger European dimension through projects of the European Commission Framework Programme. In particular, in the frame of the European Severe Accident Research NETwork of Excellence (SARNET), jointly executed research activities were performed with the ultimate objectives of providing physical models for integration into ASTEC and making the code the European reference. This effort will go on in the frame of the SARNET2 next network. The ASTEC models are today at the state of the art, except for reflooding of a degraded core. Many applications have been performed by IRSN for significant safety studies, including the probabilistic safety analysis level 2 on a French pressurized water reactor. The first version V2.0 of the new ASTEC series, released in spring 2009, will allow simulation of the European Pressurized Reactor (EPR) and will include advanced core degradation models. Then, ASTEC will remain the repository of knowledge gained from international research and development. Other long-term objectives are on one hand extension of the scope of application to boiling water reactors and CANada Deuterium Uranium (CANDU) reactors, to accidents in the ITER Fusion facility, and to Very High Temperature Reactor (VHTR) Generation IV reactors, and on the other hand to the use for emergency response tools and for severe accident simulators.

Modifications in SCDAP code for early phase degradation in a CANDU fuel channel

The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (CANDU) reactor type constitutes the major feature which prevents the direct application to this reactor design of the severe accident physical models developed for PWR/BWRs. During a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency core cooling (LOECC), particular deformation phenomena take place inside a CANDU fuel channel. Another peculiarity appears at melt relocation in horizontal geometry. Brief discussion of accident phenomena is included. SCDAP/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. The paper summarizes the applications to this reactor type previously performed by the authors with the standard (not modified) SCDAP/RELAP5 (RELAP/SCDAPSIM/MOD3.4). To obtain a code version able to simulate a CANDU channel during the early phase of a severe accident, two new models are developed inside the SCDAP routines structure. These models concern the metallic melt relocation inside the horizontal fuel channel and fuel bundle slumping. Major features of the original LIQSOL (Liquefaction-flow-Solidification) model in SCDAP are described, together with the assumptions based on experimental evidence that led to the development of the modified model. Also, two sub-models that form the basis of fuel bundle slumping treatment are presented: (a) in-built resizing of hydrodynamic sub-channels inside the fuel channel, and (b) heat transfer at direct contact, fuel to fuel and fuel to shroud. The effects of the new models on the overall performance of the code are tested in several different cases and against the results of the standard RELAP/SCDAPSIM/MOD3.4(bi7). The plans to validate the new models are discussed, as well as some general adaptation directions to different phases of a severe accident in CANDUs, foreseen for SCDAP.

조사시험용 압력용기의 조립 및 시험

Abstract The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

Electrochemical Corrosion Potential of Carbon Steel Feeder Pipes in a Canadian Deuterium Uranium Reactor Generating Station under Return-to-Service Conditions

Abstract One of the principle factors affecting the stress corrosion cracking (SCC) of a susceptible metal is its corrosion potential. A number of instances of cracking on the outlet feeder pipes o...

Numerical Calculation of Shear Stress Distribution on the Inner Wall Surface of CANDU Reactor Feeder Pipe Conveying Two-Phase Coolant

Two-phase flow fields inside feeder pipes of a CANada Deuterium Uranium (CANDU) reactor have been simulated numerically using a computational fluid dynamics (CFD) code to calculate the shear stress distribution, which is the most important factor to be considered in predicting the local areas of feeder pipes highly susceptible to flow-accelerated corrosion (FAC)-induced wall thinning. The CFD approach with schemes used in this study to simulate the turbulent flow situations inside the CANDU feeder pipes has been verified by showing a good agreement between the investigation results for the failed feedwater pipe at Surry Unit 2 plant in the U.S. and the CFD calculation. Sensitivity studies of the three geometrical parameters such as angle of the first and second bends, length of the first span between the grayloc hub and the first bend, and length of the second span between the first and second bends had already been performed. In this study, the effects of void fraction of the primary coolant coming out from the exit of pressure tubes containing nuclear fuel on the fluid shear stress distribution at the inner surface of the feeder pipe wall have been investigated to find out the local areas of feeder pipes conveying a two-phase coolant, which are highly susceptible to FAC-induced wall thinning. From the results of the CFD analysis, it is seen that the local regions of feeder pipes of the operating CANDU reactors in Korea, on which the wall thickness measurements have been performed so far, do not coincide with the worst regions predicted by the present CFD analysis, which is the connection region of straight and bend pipes near the inlet part of the bend intrados. Finally, based on the results of the present CFD analysis, a guide to the selection of the weakest local positions where the measurement of wall thickness should be performed with higher priority has been provided.

SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR–BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

Dynamic analysis of a thorium fuel cycle in CANDU reactors

The thorium fuel recycle scenarios through a Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2UO 2 and heterogeneous ThO 2UO 2–DUPIC fuels. The recycling was performed with dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, a thorium fuelled CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products for the multiple recycling fuel cycle were estimated and compared to those of a once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the spent fuel accumulation and save on the natural uranium resource when compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible than the homogeneous one. It is recommended, however, that an economic analysis be performed in the future.

Benchmark Calculation of WIMS/RFSP against Physics Measurement Data of Wolsong Nuclear Power Plants

Benchmark calculations of Canada deuterium uranium (CANDU) reactor physics design and analysis codes have been performed for a lattice code WIMS-AECL, a supercell code DRAGON, and a core analysis code RFSP by using the physics measurement data of Wolsong nuclear power plants. In this study, the lattice and reactivity device models were examined based on Wolsong-2 measurement data for the criticality and reactivity device worth. Sensitivity calculations were also performed for the number of energy groups and the cross-section library. Using the lattice and reactivity device models obtained from the Wolsong-2 calculation, the benchmark calculations were extended to the Wolsong-3 and Wolsong-4 plants. Compared to a previous study, this study showed that the results of the criticality and reactivity device worth calculations were improved when the material data were updated and the exact two-group cross sections were used. For the three nuclear power plants, the calculated core reactivity was within 0.2% Δk of criticality. The zone controller unit reactivity worth was estimated to have a maximum error of ~8%. The total reactivity worth of other reactivity control devices was consistent with the measurement data within 13%. The root-mean-square error of the flux distribution calculation was <12% when compared with flux scans performed during Phase B physics tests. In conclusion, the CANDU physics design and analysis codes used in this benchmark study predicted the physics parameters within the allowed uncertainty level of the measurement data.