Exemples d’utilisation des techniques d’optimisation en calcul de structures de reacteurs

DTIC Science & Technology

2003-03-01

34~ optimisation g~om~trique (architecture fig~e) A la difference du secteur automobile et des avionneurs, la plupart des composants des r~acteurs n...utilise des lois de comportement mat~riaux non lin~aires ainsi que des hypotheses de grands d~placements. L𔄀tude d’optimisation consiste ý minimiser...un disque simple et d~cid6 de s~lectionner trois param~tes qui influent sur la rupture : 1𔄀paisseur de la toile du disque ElI, la hauteur L3 et la

Feasible variants for intermediate storage of the spent fuel obtained at NPP Cernavoda, Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

Radu, M.; Popescu, G.

1993-12-31

The 5 CANDU-PHW Reactors of 600 Standard type of Cernavoda Nuclear Power Plant are under construction and the first unit is expected to be commissioned in 1995, group 2 following after 2 years, and then groups 3, 4 and 5 one each year. In this study there are presented feasible variants for intermediate storage of spent fuel, obtained during 30 years of operation from the stations at Cernavoda. From the solutions applied worldwide, both dry and wet storage have been taken into account. In any of the two variants, a unique intermediate storage will be provided and the storage buildingmore » was proposed to be built in 4 different stages. As a first estimation, considering the fact that, by now Romania has only one nuclear plant of CANDU fuel type the dry variant seems to be the best.« less

The Management of the Radioactive Waste Generated by Cernavoda NPP, Romania, an Example of International Cooperation - 13449

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barariu, Gheorghe

2013-07-01

The design criteria and constraints for the development of the management strategy for radioactive waste generated from operating and decommissioning of CANDU Nuclear Units from Cernavoda NPP in Romania, present many specific aspects. The main characteristics of CANDU type waste are its high concentrations of tritium and radiocarbon. Also, the existing management strategy for radioactive waste at Cernavoda NPP provides no treatment or conditioning for radioactive waste disposal. These characteristics embodied a challenging effort, in order to select a proper strategy for radioactive waste management at present, when Romania is an EU member and a signatory country of the Jointmore » Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The helping of advanced countries in radioactive waste management, directly or into the frame of the international organizations, like IAEA, become solve the aforementioned challenges at adequate level. (authors)« less

Molten salts in Nuclear Reactors (Bibliography); LES SELS FONDUS DANS LES REACTEURS NUCLEAIRES (BIBLIOGRAPHIE)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dirian, J.; Saint-James, R.

1959-01-01

A collection is presented of references dealing with the physicochemical studies of fused salts, in partictular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thoriuna are examined, and the physical properties, density, viscosity, and vapor pressure going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recovery after irradiation in a nuclear reactor is discussed. (auth)

A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bo Wook Rhee; Chang Jun Jeong; Hye Jeong Yun

2002-07-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzingmore » the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)« less

Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

NASA Astrophysics Data System (ADS)

Cook, William Gordon

Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

Tritium handling experience at Atomic Energy of Canada Limited

DOE Office of Scientific and Technical Information (OSTI.GOV)

Suppiah, S.; McCrimmon, K.; Lalonde, S.

2015-03-15

Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritiummore » powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.« less

A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lecarpentier, David; Carpentier, Vincent

2003-01-15

Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

Transformations structurales d'un acier AU Cr(9%) Mo(2%) (type ZlOCDNbV 09-02) utilisable dans les generateurs de vapeur des reacteurs a neutrons rapides

NASA Astrophysics Data System (ADS)

Vilar, Rui M.; Cizeron, Georges; Pelletier, Michel

1981-12-01

Transformations undergone by a 9 Cr-2 Mo-Nb-V steel on heating depend on the structure previously developped by quenching or tempering and on the heating rate. TTT and CCT diagrams, plotted after austenizing at 1000 and 1100°C, show only one diffusional transformation at high temperature producing equiaxed ferrite which contains a precipitate of M 23C 6 carbide; the activation energy of the process involved is 123.3 kJ/mol. At low temperatures a martensitic transformation is observed; the martensite is lath-type and autotempered.

Intergranular fracture in irradiated Inconel X-750 containing very high concentrations of helium and hydrogen

DOE Office of Scientific and Technical Information (OSTI.GOV)

Judge, Colin D.; Gauquelin, Nicolas; Walters, Lori

2015-02-01

In recent years, it has been determined that Inconel X-750 CANDU spacers have lost strength and material ductility following irradiation in reactor. The irradiated fracture behaviour of ex-service material was also found to be entirely intergranular. The heavily thermalized flux spectrum in a CANDU reactor results in transmutation of 58Ni to 59Ni. The 59Ni itself has unusually high thermal neutron reaction cross-sections of the type: (n, γ), (n, p), and (n,α). The latter two reactions, in particular, contribute to a significant enhancement of the atomic displacements in addition to creating high concentrations of hydrogen and helium within the material. Metallographicmore » examinations by transmission electron microscopy (TEM) have confirmed the presence of helium bubbles in the matrix and aligned along grain boundaries and matrix-precipitate interfaces. He bubble size and density are found to be highly dependent on the irradiation temperature and material microstructure; the bubbles are larger within grain boundary precipitates. TEM specimens extracted from fracture surfaces and crack tips give direct evidence linking crack propagation with grain boundary He bubbles.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

A statistical approach to nuclear fuel design and performance

NASA Astrophysics Data System (ADS)

Cunning, Travis Andrew

As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance, with an average sensitivity index of 48.93% on key output quantities. Pellet grain size and dish depth are also significant contributors, at 31.53% and 13.46%, respectively. A traditional limit of operating envelope case is also evaluated. This case produces output values that exceed the maximum values observed during the 105 Monte Carlo trials for all output quantities of interest. In many cases the difference between the predictions of the two methods is very prominent, and the highly conservative nature of the deterministic approach is demonstrated. A reliability analysis of CANDU fuel manufacturing parametric data, specifically pertaining to the quantification of fuel performance margins, has not been conducted previously. Key Words: CANDU, nuclear fuel, Cameco, fuel manufacturing, fuel modelling, fuel performance, fuel reliability, ELESTRES, ELOCA, dimensional reduction methods, global sensitivity analysis, deterministic safety analysis, probabilistic safety analysis.

Study of medical isotope production facility stack emissions and noble gas isotopic signature using automatic gamma-spectra analysis platform

NASA Astrophysics Data System (ADS)

Zhang, Weihua; Hoffmann, Emmy; Ungar, Kurt; Dolinar, George; Miley, Harry; Mekarski, Pawel; Schrom, Brian; Hoffman, Ian; Lawrie, Ryan; Loosz, Tom

2013-04-01

The nuclear industry emissions of the four CTBT (Comprehensive Nuclear-Test-Ban Treaty) relevant radioxenon isotopes are unavoidably detected by the IMS along with possible treaty violations. Another civil source of radioxenon emissions which contributes to the global background is radiopharmaceutical production companies. To better understand the source terms of these background emissions, a joint project between HC, ANSTO, PNNL and CRL was formed to install real-time detection systems to support 135Xe, 133Xe, 131mXe and 133mXe measurements at the ANSTO and CRL 99Mo production facility stacks as well as the CANDU (CANada Deuterium Uranium) primary coolant monitoring system at CRL. At each site, high resolution gamma spectra were collected every 15 minutes using a HPGe detector to continuously monitor a bypass feed from the stack or CANDU primary coolant system as it passed through a sampling cell. HC also conducted atmospheric monitoring for radioxenon at approximately 200 km distant from CRL. A program was written to transfer each spectrum into a text file format suitable for the automatic gamma-spectra analysis platform and then email the file to a server. Once the email was received by the server, it was automatically analysed with the gamma-spectrum software UniSampo/Shaman to perform radionuclide identification and activity calculation for a large number of gamma-spectra in a short period of time (less than 10 seconds per spectrum). The results of nuclide activity together with other spectrum parameters were saved into the Linssi database. This database contains a large amount of radionuclide information which is a valuable resource for the analysis of radionuclide distribution within the noble gas fission product emissions. The results could be useful to identify the specific mechanisms of the activity release. The isotopic signatures of the various radioxenon species can be determined as a function of release time. Comparison of 133mXe and 133Xe activity ratios showed distinct differences between the closed CANDU primary coolant system and radiopharmaceutical production releases. According to the concept proposed by Kalinowski and Pistner (2006), the relationship between different isotopic activity ratios based on three or four radioxenon isotopes was plotted in a log-log diagram for source characterisation (civil vs. nuclear test). The multiple isotopic activity ratios were distributed in three distinct areas: HC atmospheric monitoring ratios extended to far left; the CANDU primary coolant system ratios lay in the middle; and 99Mo stack monitoring ratios for ANSTO and CRL were located on the right. The closed CANDU primary coolant has the lowest logarithmic mean ratio that represents the nuclear power reactor operation. The HC atmospheric monitoring exhibited a broad range of ratios spreading over several orders of magnitude. In contrast, the ANSTO and CRL stack emissions showed the smallest range of ratios but the results indicate at least two processes involved in the 99Mo productions. Overall, most measurements were found to be shifted towards the reactor domain. The hypothesis is that this is due to an accumulation of the isotope 131mXe in the stack or atmospheric background as it has the longest half-life and extra 131mXe emissions from the decay of 131I. The contribution of older 131mXe to a fresh release shifts the ratio of 133mXe/131mXe to the left. It was also very interesting to note that there were some situations where isotopic ratios from 99Mo production emissions fell within the nuclear test domain. This is due to operational variability, such as shorter target irradiation times. Martin B. Kalinowski and Christoph Pistner, (2006), Isotopic signature of atmospheric xenon released from light water reactors, Journal of Environmental Radioactivity, 88, 215-235.

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Transmutation of actinides in power reactors.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

Tritium resources available for fusion reactors

NASA Astrophysics Data System (ADS)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future fusion reactors.

Renewability and sustainability aspects of nuclear energy

NASA Astrophysics Data System (ADS)

Şahin, Sümer

2014-09-01

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, 233U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO2/RG-PuO2) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG-PuO2 + 96 % ThO2; 6 % RG-PuO2 + 94 % ThO2; 10 % RG-PuO2 + 90 % ThO2; 20 % RG-PuO2 + 80 % ThO2; 30 % RG-PuO2 + 70 % ThO2, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ˜ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ˜ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG-PuO2 fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MWth has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ˜160 kg 233U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ˜1.3.

Tritium laboratory with multiple purposes at NIPNE Magurele Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matei, L.; Postolache, C.

2008-07-15

The Tritium Laboratory from NIPNE (Romania)) is part of Radioisotope Research and Production Center. The Tritium Laboratory has been in operation since 1960, and carries out R and D activities involving tritium sources in gaseous, liquids and solid state, provides specialized service to CANDU NPP Cernavoda (Romania)), and provides tritium assay services to internal and external customers. The paper presents the activities and perspectives of Tritium Laboratory and its performances in accordance with Quality System Management. (authors)

Modeling CANDU-6 liquid zone controllers for effects of thorium-based fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

St-Aubin, E.; Marleau, G.

2012-07-01

We use the DRAGON code to model the CANDU-6 liquid zone controllers and evaluate the effects of thorium-based fuels on their incremental cross sections and reactivity worth. We optimize both the numerical quadrature and spatial discretization for 2D cell models in order to provide accurate fuel properties for 3D liquid zone controller supercell models. We propose a low computer cost parameterized pseudo-exact 3D cluster geometries modeling approach that avoids tracking issues on small external surfaces. This methodology provides consistent incremental cross sections and reactivity worths when the thickness of the buffer region is reduced. When compared with an approximate annularmore » geometry representation of the fuel and coolant region, we observe that the cluster description of fuel bundles in the supercell models does not increase considerably the precision of the results while increasing substantially the CPU time. In addition, this comparison shows that it is imperative to finely describe the liquid zone controller geometry since it has a strong impact of the incremental cross sections. This paper also shows that liquid zone controller reactivity worth is greatly decreased in presence of thorium-based fuels compared to the reference natural uranium fuel, since the fission and the fast to thermal scattering incremental cross sections are higher for the new fuels. (authors)« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

A Hardware-in-the-Loop Simulation Platform for the Verification and Validation of Safety Control Systems

NASA Astrophysics Data System (ADS)

Rankin, Drew J.; Jiang, Jin

2011-04-01

Verification and validation (V&V) of safety control system quality and performance is required prior to installing control system hardware within nuclear power plants (NPPs). Thus, the objective of the hardware-in-the-loop (HIL) platform introduced in this paper is to verify the functionality of these safety control systems. The developed platform provides a flexible simulated testing environment which enables synchronized coupling between the real and simulated world. Within the platform, National Instruments (NI) data acquisition (DAQ) hardware provides an interface between a programmable electronic system under test (SUT) and a simulation computer. Further, NI LabVIEW resides on this remote DAQ workstation for signal conversion and routing between Ethernet and standard industrial signals as well as for user interface. The platform is applied to the testing of a simplified implementation of Canadian Deuterium Uranium (CANDU) shutdown system no. 1 (SDS1) which monitors only the steam generator level of the simulated NPP. CANDU NPP simulation is performed on a Darlington NPP desktop training simulator provided by Ontario Power Generation (OPG). Simplified SDS1 logic is implemented on an Invensys Tricon v9 programmable logic controller (PLC) to test the performance of both the safety controller and the implemented logic. Prior to HIL simulation, platform availability of over 95% is achieved for the configuration used during the V&V of the PLC. Comparison of HIL simulation results to benchmark simulations shows good operational performance of the PLC following a postulated initiating event (PIE).

Trends in the short-term release of fission products and actinides to aqueous solution from used CANDU fuels at elevated temperature

NASA Astrophysics Data System (ADS)

Stroes-Gascoyne, S.

1992-08-01

A large number of short-term leaching experiments has been performed to determine fission product and actinide release from used CANDU (CANada Deuterium Uranium) fuels and to establish which factors affect release. Results are reported after30 ± 10 d leaching at 100-150°C under oxidizing (air) or reducing (Ar-3% H 2 or Ar) conditions, in various synthetic groundwaters. Cesium-137 release (0.007-6%) was positively correlated with increases in fuel power, leachant temperature and ionic strength. Strontium-90 release (0.0003-0.3%) generally increased with ionic strength, higher temperature and redox conditions. Actinide and Tc concentrations were compared to ranges calculated with a thermodynamic equilibrium model, that accounts for the uncertain geochemical parameters of a nuclear waste vault by calculating concentration ranges based on 40000 hypothetical cases. Experimental U concentrations (10 -8.5 to 10 -3 mol/kg) were higher than the model range, probably because of higher redox potentials in the experiments. Measured Pu concentrations (10 -12.5 to 10 -7 mol/kg) were at the low end of the calculated range. Americium and Cm concentrations (10 -12.5 to 10 -7 and 10 -15 to 10 -9 mol/kg, respectively) were highest under oxidizing conditions and higher temperatures. Technetium-99 concentrations (10 -5.5 to 10 -10.5 mol/kg) covered a much narrower range than calculated by the model.

Renewability and sustainability aspects of nuclear energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Şahin, Sümer, E-mail: ssahin@atilim.edit.tr

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2};more » 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.« less

Calibration factors for the SNOOPY NP-100 neutron dosimeter

NASA Astrophysics Data System (ADS)

Moscu, D. F.; McNeill, F. E.; Chase, J.

2007-10-01

Within CANDU nuclear power facilities, only a small fraction of workers are exposed to neutron radiation. For these individuals, roughly 4.5% of the total radiation equivalent dose is the result of exposure to neutrons. When this figure is considered across all workers receiving external exposure of any kind, only 0.25% of the total radiation equivalent dose is the result of exposure to neutrons. At many facilities, the NP-100 neutron dosimeter, manufactured by Canberra Industries Incorporated, is employed in both direct and indirect dosimetry methods. Also known as "SNOOPY", these detectors undergo calibration, which results in a calibration factor relating the neutron count rate to the ambient dose equivalent rate, using a standard Am-Be neutron source. Using measurements presented in a technical note, readings from the dosimeter for six different neutron fields in six source-detector orientations were used, to determine a calibration factor for each of these sources. The calibration factor depends on the neutron energy spectrum and the radiation weighting factor to link neutron fluence to equivalent dose. Although the neutron energy spectra measured in the CANDU workplace are quite different than that of the Am-Be calibration source, the calibration factor remains constant - within acceptable limits - regardless of the neutron source used in the calibration; for the specified calibration orientation and current radiation weighting factors. However, changing the value of the radiation weighting factors would result in changes to the calibration factor. In the event of changes to the radiation weighting factors, it will be necessary to assess whether a change to the calibration process or resulting calibration factor is warranted.

Emerging nuclear programs in Asia: The Phillipines, Thailand, Indonesia, and Pakistan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, M.L.

This article is a review of the potential for nuclear energy development in the developing nations of Pakistan, Indonesia, Thailand, and the Philippines. In each country, there is a substantial need for new generating capacity, and each is exploring the idea of having nuclear energy supply a meaningful portion of this new capacity. Of the four countries, only Pakistan is currently a nuclear operator, and one vintage CANDU plant in operation and the Chashma unit under construction. Thailand and Indonesia have ambitious plans to have 12 reactors in service by the year 2015.

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

CANDU in-reactor quantitative visual-based inspection techniques

NASA Astrophysics Data System (ADS)

Rochefort, P. A.

2009-02-01

This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P.

1997-02-01

Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual inmore » detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results.« less

Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

NASA Astrophysics Data System (ADS)

Philipose, K.; Shenton, B.

2011-04-01

The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

Development of Molten Corium Using An Exothermic Chemical Reaction for the Molten- Fuel Moderator-Interaction Studies at Chalk River Laboratories

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nitheanandan, T.; Sanderson, D.B.; Kyle, G.

2004-07-01

Atomic Energy of Canada Limited (AECL) has partnered with Argonne National Laboratory to develop a corium thermite prototypical of Candu material and test the concept of ejecting {approx}25 kg of the molten material from a pressure tube with a driving pressure of 10 MPa. This development program has been completed and the technology transferred to AECL. Preparation for the molten-fuel moderator-interaction tests at AECL's Chalk River Laboratories is well underway. A mixture of 0.582 U/0.077 U{sub 3}O{sub 8}/0.151 Zr/0.19 CrO{sub 3} (wt%) as reactant chemicals has been demonstrated to produce a corium consisting of 0.73 UO{sub 2}/0.11 Zr/0.06 ZrO{sub 2}/0.10more » Cr (wt%) at {approx}2400 deg. C. This is comparable to the target Candu specific corium of 0.9 UO{sub 2}/0.1 Zr (wt%), with limited oxidation. The peak melt temperature was confirmed from small-scale thermitic reaction tests. Several small-scale tests were completed to qualify the thermite to ensure operational safety and a quantifiable experimental outcome. The proposed molten-fuel moderator-interaction experiments at Chalk River Laboratories will consist of heating the thermite mixture inside a 1.14-m long insulated pressure tube. Once the molten material has reached the desired temperature of {approx}2400 deg. C, the pressure inside the tube will be raised to about 10 MPa, and the pressure tube will fail at a pre-machined flaw, ejecting the molten material into the surrounding tank of water. The test apparatus, instrumentation, data acquisition and control systems have been assembled, and a series of successful commissioning tests have been completed. (authors)« less

An Example of an INPRO Assessment of an INS in the Area of Waste Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allan, C.; Busurin, Y.; Depisch, F.

2006-07-01

Following a resolution of the General Conference of the IAEA in the year 2000 the International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contributionmore » to meeting the energy needs of the 21. century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. The manuals set out the information that needs to be assembled to perform an assessment and provide guidance on selecting the acceptance limits and, for a given INS, for determining the value of the indicators for comparison with the associated acceptance limits. Each manual also includes an example of a specific assessment to illustrate the guidance. This paper discusses the example presented in the manual for performing an INPRO assessment in the area of waste management. The example, chosen solely for the purpose of illustrating the INPRO methodology, describes an assessment of an INS based on the DUPIC fuel cycle. It is assumed that uranium is mined, milled, converted, enriched, and fabricated into LWR fuel in Canada. The LWR fuel is assumed to be leased to a utility in the USA. The spent LWR fuel is assumed to be returned to Canada where it is processed into CANDU DUPIC fuel, which is then burned in CANDU reactors. The assessment steps and the results are presented in detail in the paper. The example illustrates an assessment performed for an INS at an early stage of development. (authors)« less

The Results From the First High-Pressure Melt Ejection Test Completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nitheanandan, T.; Kyle, G.; O'Connor, R.

2006-07-01

A high-pressure melt ejection test using prototypical corium was conducted at Atomic Energy of Canada Limited Chalk River Laboratories. This test was planned by the CANDU Owners Group to study the potential for an energetic interaction between molten fuel and water under postulated single-channel flow-blockage events. The experiments were designed to address regulator concerns surrounding this very low probability postulated accident events in CANDU Pressurized Heavy Water Reactors. The objective of the experimental program is to determine whether a highly energetic 'steam explosion' and associated high-pressure pulse, is possible when molten material is finely fragmented as it is ejected frommore » a fuel channel into the heavy-water moderator. The finely fragmented melt particles would transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection test consisted of heating up a {approx} 5 kg thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3} inside a 1.14-m length of insulated pressure tube. When the molten material reached the desired temperature of {approx} 2400 deg C, the pressure inside the tube was raised to 11.6 MPa, failing the pressure tube at a pre-machined flaw, and releasing the molten material into the surrounding tank of 68 deg C water. The experiment investigated the dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels. The measured mean particle size was 0.686 mm and the peak dynamic pressures were between 2.54 and 4.36 MPa, indicating that an energetic interaction between the melt and the water did not occur in the test. (authors)« less

Faith in public power paid off for Ontario

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smart, R.J.

Ontario Hydro is the result of a grassroots belief that public power would create an industrial tax base and benefit the entire community. After 75 years, Ontario Hydro has $18 billion in assets, a dependable 22.6 thousand-MW peak capacity, and residential utility bills that are only one third of New York City's. The potential of Niagara Falls as a power source prompted a 1902 drive for public power which took another 19 years to realize. An overview of Ontario Hydro's development highlights the role of a 25-man syndicate led by Adam Beck, the influence of Ontario Hydro on the developersmore » of the Tennessee Valley Authority, the tapping of the St. Lawrence River, and the CANDU nuclear program. 1 figure. (DCK)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katchadjian, Pablo, E-mail: katcha@cnea.gov.ar; Desimone, Carlos, E-mail: katcha@cnea.gov.ar; Garcia, Alejandro, E-mail: katcha@cnea.gov.ar

Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuringmore » wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces.« less

Design of Test Loops for Forced Convection Heat Transfer Studies at Supercritical State

NASA Astrophysics Data System (ADS)

Balouch, Masih N.

Worldwide research is being conducted to improve the efficiency of nuclear power plants by using supercritical water (SCW) as the working fluid. One such SCW reactor considered for future development is the CANDU-Supercritical Water Reactor (CANDU-SCWR). For safe and accurate design of the CANDU-SCWR, a detailed knowledge of forced-convection heat transfer in SCW is required. For this purpose, two supercritical fluid loops, i.e. a SCW loop and an R-134a loop are developed at Carleton University. The SCW loop is designed to operate at pressures as high as 28 MPa, temperatures up to 600 °C and mass fluxes of up to 3000 kg/m2s. The R-134a loop is designed to operate at pressures as high as 6 MPa, temperatures up to 140 °C and mass fluxes in the range of 500-6000 kg/m2s. The test loops designs allow for up to 300 kW of heating power to be imparted to the fluid. Both test loops are of the closed-loop design, where flow circulation is achieved by a centrifugal pump in the SCW loop and three parallel-connected gear pumps in the R-134a loop, respectively. The test loops are pressurized using a high-pressure nitrogen cylinder and accumulator assembly, which allows independent control of the pressure, while simultaneously dampening pump induced pressure fluctuations. Heat exchangers located upstream of the pumps control the fluid temperature in the test loops. Strategically located measuring instrumentation provides information on the flow rate, pressure and temperature in the test loops. The test loops have been designed to accommodate a variety of test-section geometries, ranging from a straight circular tube to a seven-rod bundle, achieving heat fluxes up to 2.5 MW/m2 depending on the test-section geometry. The design of both test loops allows for easy reconfiguration of the test-section orientation relative to the gravitational direction. All the test sections are of the directly-heated design, where electric current passing through the pressure retaining walls of the test sections provides the Joule heating required to heat up the fluid to supercritical conditions. A high-temperature dielectric gasket isolates the current carrying parts of the test section from the rest of the assembly. Temperature and pressure drop data are collected at the inlet and outlet, and along the heated length of the test section. The test loops and test sections are designed according to American Society of Mechanical Engineers (ASME) Pressure Piping B31.1, and Boiler and Pressure Vessel Code, Section VIII-Division 1 rules. The final test loops and test sections assemblies are certified by Technical Standards and Safety Authority (TSSA). Every attempt is made to use off-the-shelf components where possible in order to streamline the design process and reduce costs. Following a rigorous selection process, stainless steel Types 316 and 316H are selected as the construction materials for the test loops, and Inconel 625 is selected as the construction material for the test sections. This thesis describes the design of the SCW and R-134a loops along with the three test-section geometries (i.e., tubular, annular and bundle designs).

Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

NASA Astrophysics Data System (ADS)

Adenariwo, Adepoju

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.

La détection infrarouge avec les plans focaux non refroidis : état de l'artUncooled focal plane infrared detectors: the state of the art

NASA Astrophysics Data System (ADS)

Tissot, Jean-Luc

2003-12-01

The emergence of uncooled detectors has opened new opportunities for IR detection for both military and commercial applications. Development of such devices involves a lot of trade-offs between the different parameters that define the technological stack. These trade-offs explain the number of different architectures that are under worldwide development. The key factor is to find a high sensitivity and low noise thermometer material compatible with silicon technology in order to achieve high thermal isolation in the smallest area as possible. Ferroelectric thermometer based hybrid technology and electrical resistive thermometer based (microbolometer) technology are under development. LETI and ULIS have chosen from the very beginning to develop first a monolithic microbolometer technology fully compatible with commercially available CMOS technology and secondly amorphous silicon based thermometer. This silicon approach has the greatest potential for reducing infrared detector manufacturing cost. After the development of the technology, the transfer to industrial facilities has been performed in a short period of time and the production is now ramping up with ULIS team in new facilities. LETI and ULIS are now working to facilitate the IRFPA integration into equipment in order to address a very large market. Achievement of this goal needs the development of smart sensors with on-chip advanced functions and the decrease of manufacturing cost of IRFPA by decreasing the pixel pitch and simplifying the vacuum package. We present in this paper the technology developed by CEA/LETI and its improvement for being able to designs 384×288 and 160×120 arrays with a pitch of 35 μm. Thermographic application needs high stability infrared detector with a precise determination of the amount of absorbed infrared flux. Hence, infrared detector with internal temperature stabilized shield has been developed and characterized. These results will be presented. To cite this article: J.-L. Tissot, C. R. Physique 4 (2003).

View of Pakistan Atomic Energy Commission towards SMPR's in the light of KANUPP performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huseini, S.D.

1985-01-01

The developing countries in general do not have grid capacities adequate enough to incorporate standard size, economic but rather large nuclear power plants for maximum advantage. Therefore, small and medium size reactors (SMPR) have been and still are, of particular interest to the developing countries in spite of certain known problems with these reactors. Pakistan Atomic Energy Commission (PAEC) has been operating a CANDU type of a small PHWR plant since 1971 when it was connected to the local Karachi grid. This paper describes PAEC's view in the light of KANUPP performance with respect to such factors associated with SMPR'smore » as selection of suitable reactor size and type, its operation in a grid of small capacity, flexibility of operation and its role as a reliable source of electrical power.« less

Automatic ultrasonic inspection system for wear determination in calandria tubes of Embalse Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Katchadjian, Pablo; Desimone, Carlos; Garcia, Alejandro; Antonaccio, Carlos; Schroeter, Fernando; Molina, Héctor

2015-03-01

Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuring wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces.

Methodology, status and plans for development and assessment of TUF and CATHENA codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luxat, J.C.; Liu, W.S.; Leung, R.K.

1997-07-01

An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically onmore » CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.« less

Plant maintenance and plant life extension issue, 2008

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agnihotri, Newal

The focus of the March-April issue is on plant maintenance and plant life extension. Major articles include the following: Exciting time to be at the U.S. NRC, by Dale Klein, Nuclear Regulatory Commission; Extraordinary steps to ensure a minimal environmental impact, by George Vanderheyden, UniStar Nuclear Energy, LLC.; Focused on consistent reduction of outages, by Kevin Walsh, GE Hitachi Nuclear Energy; On the path towards operational excellence, by Ricardo Perez, Westinghouse Electric Company; Ability to be refuelled on-line, by Ian Trotman, CANDU Services, Atomic Energy of Canada, Ltd.; ASCA Application for maintenance of SG secondary side, by Patrick Wagner, Wolfmore » Creek Nuclear Operating Corporation, Phillip Battaglia and David Selfridge, Westinghouse Electric Company; and, An integral part of the landscape and lives, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Steam generator bowl drain repairs, by John Makar and Richard Gimple, Wolf Creek Nuclear Operating Corporation.« less

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

Presence de Carbone-13 dans les elements combustibles de type (U,Pu)O 2 irradies en reacteur rapide

NASA Astrophysics Data System (ADS)

Kryger, Bernard; Hagemann, Robert

1982-06-01

Du carbone-13 produit par la réaction de capture neutronique 168O + 10n → 136C + 42He se forme dans les combustibles de type oxyde irradiés en neutrons rapides. Cette réaction, dont le seuil d'énergie se situe à 2.35 MeV, conduit à la formation d'une quantité de carbone-13 qui peut varier notablement suivant le spectre neutronique du réacteur (entre 20 et 40 × 10 -6g 13C/g (U,Pu)O 2 pour une fluence de 2 × 10 23 n/cm 2). DES mesures effectuées sur le combustible et la gaine par spectrométrie de masse après irradiation montrent qu'une fraction égale ou supérieure à la moitié du carbone-13 produit dans l'oxyde peut être transférée dans la gaine. Un tel comportement nous fait considérer le carbone-13 comme un véritable marqueur du carbone plus généralement contenu dans l'oxyde et, à ce titre, la détection de cet isotope devrait contribuer à élucider tout particulièrement les mécanismes de carburation de la gaine par les combustibles (U,Pu)O 2 des réacteurs surgénérateurs.

Improvement of Pt/C/PTFE catalyst type used for hydrogen isotope separation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vasut, F.; Preda, A.; Zamfirache, M.

2008-07-15

The CANDU reactor from the Nuclear Power plant Cernavoda (Romania)) is the most powerful tritium source from Europe. This reactor is moderated and cooled by heavy water that becomes continuously contaminated with tritium. Because of this reason, the National R and amp;D Inst. for Cryogenic and Isotopic Technologies developed a detritiation technology based on catalytic isotopic exchange and cryogenic distillation. The main effort of our Inst. was focused on finding more efficient catalysts with a longer operational life. Some of the tritium removal processes involved in Fusion Science and Technology use this type of catalyst 1. Several Pt/C/PTFE hydrophobic catalystsmore » that could be used in isotopic exchange process 2,3,4 were produced. The present paper presents a comparative study between the physical and morphological properties of different catalysts manufactured by impregnation at our institute. The comparison consists of a survey of specific surface, pores volume and pores distribution. (authors)« less

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Wavelet analysis of poorly-focused ultrasonic signal of pressure tube inspection in nuclear industry

NASA Astrophysics Data System (ADS)

Zhao, Huan; Gachagan, Anthony; Dobie, Gordon; Lardner, Timothy

2018-04-01

Pressure tube fabrication and installment challenges combined with natural sagging over time can produce issues with probe alignment for pressure tube inspection of the primary circuit of CANDU reactors. The ability to extract accurate defect depth information from poorly focused ultrasonic signals would reduce additional inspection procedures, which leads to a significant time and cost saving. Currently, the defect depth measurement protocol is to simply calculate the time difference between the peaks of the echo signals from the tube surface and the defect from a single element probe focused at the back-wall depth. When alignment issues are present, incorrect focusing results in interference within the returning echo signal. This paper proposes a novel wavelet analysis method that employs the Haar wavelet to decompose the original poorly focused A-scan signal and reconstruct detailed information based on a selected high frequency component range within the bandwidth of the transducer. Compared to the original signal, the wavelet analysis method provides additional characteristic defect information and an improved estimate of defect depth with errors less than 5%.

Fissuration d'un composite a matrice ceramique dans une application de piece statique de turbine a gaz

NASA Astrophysics Data System (ADS)

Lebel, Larry

Une procedure experimentale a ete developpee pour caracteriser les mecanismes de deterioration et la durabilite de materiaux composites a matrice ceramique (CMC) dans une application de piece statique de turbine a gaz. Tandis que la plupart des essais de caracterisation publies sur les materiaux CMC ont ete realises sous des conditions de chargement controle, la presente recherche tente de reproduire la relaxation des contraintes qui se produit normalement dans une piece statique a haute temperature. Dans l'experience proposee, un echantillon planaire de forme haltere est chauffe de facon cyclique sur une de ses faces et refroidi sur l'autre, tout en etant contraint dans ses deplacements. La contrainte de flexion resultante au centre de l'echantillon, mesuree par une cellule de charge, correspond a la contrainte de flexion qui a ete prealablement predite au centre des panneaux d'une chambre a combustion generique. Un materiau CMC multicouche compose d'une matrice d'alumine poreuse et de fibres NextelMD 720 a ete utilise pour developper l'experience. Des essais de calibration ont d'abord ete realises en utilisant un systeme de chauffage par lampe infrarouge, atteignant jusqu'a 1160 °C a la surface de l'echantillon. Un systeme laser au CO2 a par la suite ete utilise pour realiser des essais de deterioration a haute puissance, atteignant en fin d'essai des temperatures de surface excedant la limite de 1200 °C du materiau et des differences de temperature a travers l'epaisseur de plus de 1000 °C. Sous la puissance de chauffage imposee a amplitude constante, l'accumulation de dommage a fait en sorte d'augmenter la temperature en surface et les gradients de temperature a travers le materiau. Une reduction de la contrainte dans le temps a ete observee a cause du fluage, de la fissuration et de la delamination du materiau sous la condition de confinement du deplacement, menant a une stabilisation du niveau de dommage a une certaine profondeur dependant de la contrainte thermique initiale. La procedure de caracterisation developpee s'avere etre un outil prometteur pour developper de nouveaux types de materiaux, de meme que pour comparer la durabilite de materiaux existants sous des conditions representatives de pieces statiques de turbine a gaz. None

Development of a Korean reference HLW disposal system under the Korean representative geologic conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Heui-Joo; Lee, Jong Youl; Choi, Jongwon

2007-07-01

The development of a Korean Reference disposal System for the spent fuels from PWR and CANDU reactors is outlined in this paper. Around 36,000 tU of spent fuels are being projected based on the lifetimes of 28 nuclear power reactors in Korea. Since the site for the geological disposal has not yet been decided, a hypothetical site with representative Korean geologic conditions is proposed for the conceptual design of the repository. The disposal rates of the spent fuels are determined according to the total operation time of 55 years. The canisters are optimized by considering natural Korean conditions, and themore » buffer is designed with domestic Ca-bentonite. The depth of the repository is determined to be 500 m below the ground's surface. The canister separation distances are determined through a thermal analysis. The main features of the repository are presented from the layout to the closure. A computer program has been developed to calculate and analyze the volume and the area of the disposal system to help in the cost analysis. The final output of the design is presented as a unit disposal cost, US $315 /kgU. (authors)« less

Investigation related to hydrogen isotopes separation by cryogenic distillation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bornea, A.; Zamfirache, M.; Stefanescu, I.

2008-07-15

Research conducted in the last fifty years has shown that one of the most efficient techniques of removing tritium from the heavy water used as moderator and coolant in CANDU reactors (as that operated at Cernavoda (Romania)) is hydrogen cryogenic distillation. Designing and implementing the concept of cryogenic distillation columns require experiments to be conducted as well as computer simulations. Particularly, computer simulations are of great importance when designing and evaluating the performances of a column or a series of columns. Experimental data collected from laboratory work will be used as input for computer simulations run at larger scale (formore » The Pilot Plant for Tritium and Deuterium Separation) in order to increase the confidence in the simulated results. Studies carried out were focused on the following: - Quantitative analyses of important parameters such as the number of theoretical plates, inlet area, reflux flow, flow-rates extraction, working pressure, etc. - Columns connected in series in such a way to fulfil the separation requirements. Experiments were carried out on a laboratory-scale installation to investigate the performance of contact elements with continuous packing. The packing was manufactured in our institute. (authors)« less

Pourbaix Diagrams at Elevated Temperatures A Study of Zinc and Tin

NASA Astrophysics Data System (ADS)

Palazhchenko, Olga

Metals in industrial settings such as power plants are often subjected to high temperature and pressure aqueous environments, where failure to control corrosion compromises worker and environment safety. For instance, zircaloy (1.2-1.7 wt.% Sn) fuel rods are exposed to aqueous 250-310 °C coolant in CANDU reactors. The Pourbaix (EH-pH) diagram is a plot of electrochemical potential versus pH, which shows the domains of various metal species and by inference, corrosion susceptibility. Elevated temperature data for tin +II and tin +IV species were obtained using solid-aqueous phase equilibria with the respective oxides, in a batch vessel with in-situ pH measurement. Solubilities, determined via spectroscopic techniques, were used to calculate equilibrium constants and the Gibbs energies of Sn complexes for E-pH diagram construction. The SnOH3+ and Sn(OH )-5 species were incorporated, for the first time, into the 298.15 K and 358.15 K diagrams, with novel Go values determined at 358.15 K. Key words: Pourbaix diagrams, EH-pH, elevated temperatures, solubility, equilibrium, metal oxides, hydrolysis, redox potential, pH, thermochemical data, tin, zinc, zircaloy, corrosion, passivity.

Enrichment of 57Fe isotope in neutron flux of nuclear reactors observed by Mössbauer spectroscopy.

PubMed

Sawicki, Jerzy A

2018-02-01

The abundance of 57 Fe isotope in nuclear reactor core materials can be considerably enriched by neutron-capture 56 Fe(n,γ) reactions. This is demonstrated using the sections of Zr-2.5 wt.%Nb pressure tubes removed from two CANDU* reactors. The tubes contained 0.11 and 0.04wt% Fe and were irradiated for about 10 effective full power years (EFPY) up to ~10 26 n/m 2 fast neutron (E > 1MeV) fluencies. The Mössbauer spectra of 57 Fe in irradiated samples indicated up to 10 times larger areas than unirradiated off-cuts from the same pressure tubes. The observed effect is in accord with the values calculated for known thermal neutron-capture cross-sections and resonance capture integrals, neutron flux profiles and spectra, and times of irradiation. The build-up of 57 Fe facilitated recording Mössbauer absorption spectra of alloys with minor amount of Fe down to ~ 400ppm, despite intense background radiation emitted by samples. These findings can open new possibilities in post-irradiation studies of alloys used in nuclear reactors and in other objects subjected to large neutron fluencies. Copyright © 2017 Elsevier Ltd. All rights reserved.

Localization of a Robotic Crawler for CANDU Fuel Channel Inspection

NASA Astrophysics Data System (ADS)

Manning, Mark

This thesis discusses the design and development of a pipe crawling robot for the purpose of CANDU fuel channel inspection. The pipe crawling robot shall be capable of deploying the existing CIGAR (Channel Inspection and Gauging Apparatus for Reactors) sensor head. The main focus of this thesis is the design of the localization system for this robot and the many tests that were completed to demonstrate its accuracy. The proposed localization system consists of three redundant resolver wheels mounted to the robot's frame and two resolvers that are mounted inside a custom made cable drum. This cable drum shall be referred to in this thesis as the emergency retrieval device. This device serves the dual-purpose of providing absolute position measurements (via the cable that is tethered to the robot) as well as retrieving the robot if it is inoperable. The estimated accuracy of the proposed design is demonstrated with the use of a proof-of-concept prototype and a custom made test bench that uses a vision system to provide a more accurate estimate of the robot's position. The only major difference between the proof-of-concept prototype and the proposed solution is that the more expensive radiation hardened components were not used in the proof-of-concept prototype design. For example, the proposed solution shall use radiation hardened resolver wheels, whereas the proof-of-concept prototype used encoder wheels. These encoder wheels provide the same specified accuracy as the radiation hardened resolvers for the most realistic results possible. The rationale behind the design of the proof-of-concept prototype, the proposed final design, the design of the localization system test bench, and the test plan for developing all of the components of the design related to the robot's localization system are discussed in the thesis. The test plan provides a step by step guide to the configuration and optimization of an Unscented Kalman Filter (UKF). The UKF was selected as the ideal sensor fusion algorithm for use in this application. Benchmarking was completed to compare the accuracy achieved by the UKF algorithm to other data fusion algorithms. When compared to other algorithms, the UKF demonstrated the best accuracy when considering all likely sources of error such as sensor failure and surface unevenness. The test results show that the localization system is able to achieve a worst case positional accuracy of +/- 3.6 mm for the robot crawler over the full 6350 mm distance that the robot travels inside the pressure tube. This is extrapolated from the test results completed over the shorter length test bench with simulated surface unevenness. The key benefits of the pipe crawling robot when compared to the current system include: reduced dosage to workers and the reduced outage time. The advantages are due to the fact that the robot can be automated and multiple inspection robots can be deployed simultaneously. The current inspection system is only able to complete one inspection at a time.

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Conception et calibration d'un sonoreacteur pour l'oxydation de la cellulose par le systeme TEMPO/NaOCl/NaBr

NASA Astrophysics Data System (ADS)

Paquin, Michel

Avec le contexte economique actuel dans le domaine des pates et papiers au Canada, l'industrie se doit de diversifier ses produits mis en marche. La fermeture de plus de 20 usines depuis 2005, une baisse du PIB de l'industrie de 1,4 milliard CAD entre 1999--2008, une baisse de la demande de 2,4 %, une diminution du prix de la pate de 20,9 % depuis juillet 2009. La delocalisation du secteur vers l'Asie et l'hemisphere sud sont autant de raisons pour laquelle l'industrie se doit d'etre a l'avant plan de nouvelle technologie a base de fibre de bois. Pour augmenter leur rentabilite, l'industrie se doit de diversifier ses produits dans d'autres secteurs que le simple fabricant de papier impression-ecriture. Sa diversification passe par l'elaboration de nouveaux papiers a valeur ajoutee (papier conducteur, papier bioactif, etc.), par l'utilisation de la biomasse forestiere pour la production d'energie, par l'utilisation de la biomasse forestiere pour l'elaboration d'une plateforme de chimie verte, par l'utilisation de la lignine pour le developpement de polymeres et par l'utilisation de la fibre cellulosique pour la fabrication de nanomateriaux. La fabrication de nanofibrille de cellulose peut devenir un des produits qui servira a diversifier la production des usines de pates et papiers. Les nanofibrilles de cellulose possedent des proprietes mecaniques et chimiques exceptionnelles. Les nanofibrilles de cellulose sont fabriquees a partir d'une oxydation selective de la pate kraft de feuillu avec le systeme TEMPO-NaOCl-NaBr. L'oxydation selective de l'alcool primaire en C6 du monomere de glucose sous forme de carboxylates engendre une modification chimique de la cellulose qui accroit l'hydrophilicite des fibrilles. Suite a cette oxydation, nous devons effectuer une desintegration mecanique de la fibre kraft de feuillu oxydee pour separer les fibrilles. Le processus d'oxydation de la fibre par le systeme TEMPO-NaOCl-NaBr et sa defibrillation par la suite engendre une grande consommation d'energie et de reactif qui rend le procede difficilement accessible au milieu industriel. L'utilisation des ultrasons lors de la reaction d'oxydation permet de reduire de 50% le temps de reaction et d'autant la consommation de produits chimiques. Actuellement le processus d'oxydation s'effectue en mode discontinu pour une quantite de 20 grammes de pate dans un reacteur en verre dans un bain a ultrason. L'objectif principal de ce travail est d'elaborer un sonoreacteur en mode semi-continu afin d'etre en mesure de transferer une reaction d'oxydation du mode discontinu en mode semi-continu. Le transfert de reaction sera effectue en realisant la calibration de l'activite acoustique des deux differents reacteurs, discontinu et semi-continu, par la methode de Weissler. La methode de Weissler quantifie le taux de production radicalaire en mesurant la formation d'iode moleculaire d'une solution aqueuse d'iodure de potassium sous ultrason. Suite a sa calibration, le sonoreacteur pilote (mode semi-continu) a demontre une augmentation de la production radicalaire de 683 % a 170 kHz en comparaison avec l'utilisation d'un bain a ultrason a la meme frequence. Lors de la reaction d'oxydation, la puissance optimale utilisee dans le bain a ultrason a 170 kHz est de 1000 W. La puissance utilisee selon les resultats de calibration obtenue dans le sonoreacteur pilote est de 125 W soit une diminution de l'energie appliquee de 87,5 %. Lors de la reaction d'oxydation, le taux de production des groupements carboxylates est de 2,6 mmol COOH kg/min dans le bain a ultrason et de 6,87 mmol COOH kg/min dans le sonoreacteur pilote soit une augmentation de 164% du taux de formation. Selon les resultats obtenus, l'utilisation d'un sonoreacteur en mode continu peut etre envisageable pour la reaction d'oxydation de la cellulose par le systeme TEMPO-NaOCl-NaBr en industrie au niveau industriel.

ORCHID - a computer simulation of the reliability of an NDE inspection system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moles, M.D.C.

1987-03-01

CANDU pressurized heavy water reactors contain several hundred horizontally-mounted zirconium alloy pressure tubes. Following a pressure tube failure, a pressure tube inspection system called CIGARette was rapidly designed, manufactured and put in operation. Defects called hydride blisters were found to be the cause of the failure, and were detected using a combination of eddy current and ultrasonic scans. A number of improvements were made to CIGARette during the inspection period. The ORCHID computer program models the operation of the delivery system, eddy current and ultrasonic systems by imitating the on-reactor decision-making procedure. ORCHID predicts that during the early stage ofmore » development, less than one blistered tube in three would be detected, while less than one in two would be detected in the middle development stage. However, ORCHID predicts that during the late development stage, probability of detection will be over 90%, primarily due to the inclusion of axial ultrasonic scans (a procedural modification). Rotational and axial slip could severely reduce probability of detection. Comparison of CIGARette's inspection data with ORCHID's predictions indicate that the latter are compatible with the actual inspection results, through the numbers are small and data uncertain. It should be emphasized that the CIGARette system has been essentially replaced with the much more reliable CIGAR system.« less

XPS studies of UO 2 oxidation by alpha radiolysis of water at 100°C

NASA Astrophysics Data System (ADS)

Sunder, S.; Boyer, G. D.; Miller, N. H.

1990-12-01

The effect of alpha radiolysis of water on the oxidation and dissolution of UO 2 was studied at 100°C as a function of alpha-field strength and water chemistry using X-ray photoelectron spectroscopy. In N 2-purged solutions the oxidation of UO 2 increases with the strength of the alpha flux; an alpha flux greater than or equal to that from a 250-μ Ci americium-241 source leads to oxidation of UO 2 beyond the UO 2.33 (U 3O 7) stage, and an alpha flux equal to that from a 5-μ Ci source does not result in UO 2 oxidation beyond the UO 2.33 stage. The presence of dissolved H 2 in water, at a concentration ≥ 1.6 × 10 -4moldm-3, reduces the oxidation and dissolution of UO 2 due to alpha radiolysis at temperatures ≥ 100° C. It is concluded that the radiolysis of groundwater at ~ 100°C, due to the alpha flux associated with used CANDU fuel, is unlikely to make a significant contribution to its oxidative dissolution in the geological disposal vault planned in the Canadian Nuclear Fuel Waste Management Program. CANada Deuterium Uranium. Registered trademark.

Development of a tritium recovery system from CANDU tritium removal facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Draghia, M.; Pasca, G.; Porcariu, F.

2015-03-15

The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consistsmore » of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)« less

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

WTEC monograph on instrumentation, control and safety systems of Canadian nuclear facilities

NASA Technical Reports Server (NTRS)

Uhrig, Robert E.; Carter, Richard J.

1993-01-01

This report updates a 1989-90 survey of advanced instrumentation and controls (I&C) technologies and associated human factors issues in the U.S. and Canadian nuclear industries carried out by a team from Oak Ridge National Laboratory (Carter and Uhrig 1990). The authors found that the most advanced I&C systems are in the Canadian CANDU plants, where the newest plant (Darlington) has digital systems in almost 100 percent of its control systems and in over 70 percent of its plant protection system. Increased emphasis on human factors and cognitive science in modern control rooms has resulted in a reduced workload for the operators and the elimination of many human errors. Automation implemented through digital instrumentation and control is effectively changing the role of the operator to that of a systems manager. The hypothesis that properly introducing digital systems increases safety is supported by the Canadian experience. The performance of these digital systems has been achieved using appropriate quality assurance programs for both hardware and software development. Recent regulatory authority review of the development of safety-critical software has resulted in the creation of isolated software modules with well defined interfaces and more formal structure in the software generation. The ability of digital systems to detect impending failures and initiate a fail-safe action is a significant safety issue that should be of special interest to nuclear utilities and regulatory authorities around the world.

The environmental constraint needs for design improvements to the Saligny I/LLW-repository near Cernavoda NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barariu, Gheorghe

2007-07-01

The paper presents the new perspectives on the development of the L/ILW Final Repository Project which will be built near Cernavoda NPP. The Repository is designed to satisfy the main performance objectives in accordance to IAEA recommendation. Starting in October 1996, Romania became a country with an operating nuclear power plant. Reactor 2 reached the criticality on May 6, 2007 and it will be put in commercial operation in September 2007. The Ministry of Economy and Finance has decided to proceed with the commissioning of Units 3 and 4 of Cernavoda NPP till 2014. The Strategy for radioactive waste managementmore » was elaborated by National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for definitive disposal and the coordination of nuclear spent fuel and radioactive waste management (Order 844/2004) with attributions established by Governmental Decision (GO) 31/2006. The Strategy specifies the commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP in 2014. When designing the L/IL Radwaste Repository, the following prerequisites have been taken into account: 1) Cernavoda NPP will be equipped with 4 Candu 6 units. 2) National Legislation in radwaste management will be reviewed and/or completed to harmonize with UE standards 3) The selected site is now in process of confirmation after a comprehensive set of interdisciplinary investigations. (author)« less

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Iron Silicide Formation by Precipitation in a Silicon Bicrystal

NASA Astrophysics Data System (ADS)

Portier, X.; Ihlal, A.; Rizk, R.

1997-05-01

Segregation and precipitation of iron in a = 25 silicon bicrystal have been carefully investigated by means of high resolution electron microscopy and energy dispersive X-ray analyses, in combination with capacitance and electron beam induced current measurements. After intentional incorporation of iron in the bicrystal by a simple heating procedure, it was shown that a non-equilibrium segregation of iron has occurred after rapid cooling whereas iron precipitates have been produced upon slow cooling. The silicides are formed mainly at the grain boundary area and they were found to belong to the -FeSi cubic or -FeSi2 tetragonal phases. Each precipitate is simply oriented with respect to one of the two grains without any preference between them. The orientation relationships were found in perfect agreement with those observed for the corresponding iron silicides that are epitaxially grown on oriented silicon substrates. Barrier and recombinative effects on the contaminated (1200 °C) and slowly cooled samples have been detected. These effects have been associated with the formation of iron silicides at the grain boundary. La ségrégation ainsi que la précipitation de siliciures de fer au joint de grains = 25 de silicium ont été etudiées en utilisant la dispersion d'énergie des électrons, la microscopie électronique en transmission haute résolution ainsi que des mesures électriques capacitives et des mesures de courants induits par faisceau d'électrons. A la suite d'une contamination volontaire par diffusion thermique du fer au sein du bicristal, nous avons montré qu'une ségrégation hors-équilibre d'atomes de fer est obtenue après un refroidissement rapide alors qu'un refroidissement lent a pour conséquence la formation de siliciures de fer. Ces petits cristaux de siliciures croissent de préférence au niveau du joint de grains et ils ont pour phase, la phase cubique -FeSi ou la phase quadratique α-FeSi2. Chaque précipité est orienté simplement par rapport à l'un ou à l'autre des deux grains et leurs relations d'orientation coincident avec celles observées pour ces mêmes siliciures épitaxiés sur des surfaces de silicium. Les échantillons contaminés (1200 °C) et refroidis lentement présentent des barrières de potentiel et des effets recombinants. Ces activités électriques ont été associées à la présence de siliciures au niveau du joint.

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science – Addressing the ODWAC New Recommendation

PubMed Central

Dingwall, S.; Mills, C.E.; Phan, N.; Taylor, K.; Boreham, D.R.

2011-01-01

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered. The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health. Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective. PMID:21431084

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science - Addressing the ODWAC New Recommendation.

PubMed

Dingwall, S; Mills, C E; Phan, N; Taylor, K; Boreham, D R

2011-02-22

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered.The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health.Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

Current Status of The Romanian National Deep Geological Repository Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Radu, M.; Nicolae, R.; Nicolae, D.

2008-07-01

Construction of a deep geological repository is a very demanding and costly task. By now, countries that have Candu reactors, have not processed the spent fuel passing to the interim storage as a preliminary step of final disposal within the nuclear fuel cycle back-end. Romania, in comparison to other nations, represents a rather small territory, with high population density, wherein the geological formation areas with radioactive waste storage potential are limited and restricted not only from the point of view of the selection criteria due to the rocks natural characteristics, but also from the point of view of their involvementmore » in social and economical activities. In the framework of the national R and D Programs, series of 'Map investigations' have been made regarding the selection and preliminary characterization of the host geological formation for the nation's spent fuel deep geological repository. The fact that Romania has many deposits of natural gas, oil, ore and geothermal water, and intensively utilizes soil and also is very forested, cause some of the apparent acceptable sites to be rejected in the subsequent analysis. Currently, according to the Law on the spent fuel and radioactive waste management, including disposal, The National Agency of Radioactive Waste is responsible and coordinates the national strategy in the field and, subsequently, further actions will be decided. The Romanian National Strategy, approved in 2004, projects the operation of a deep geological repository to begin in 2055. (authors)« less

Comprehensive characterization of measurement data gathered by the pressure tube to calandria tube gap probe

NASA Astrophysics Data System (ADS)

Shokralla, Shaddy Samir Zaki

Multi-frequency eddy current measurements are employed in estimating pressure tube (PT) to calandria tube (CT) gap in CANDU fuel channels, a critical inspection activity required to ensure fitness for service of fuel channels. In this thesis, a comprehensive characterization of eddy current gap data is laid out, in order to extract further information on fuel channel condition, and to identify generalized applications for multi-frequency eddy current data. A surface profiling technique, generalizable to multiple probe and conductive material configurations has been developed. This technique has allowed for identification of various pressure tube artefacts, has been independently validated (using ultrasonic measurements), and has been deployed and commissioned at Ontario Power Generation. Dodd and Deeds solutions to the electromagnetic boundary value problem associated with the PT to CT gap probe configuration were experimentally validated for amplitude response to changes in gap. Using the validated Dodd and Deeds solutions, principal components analysis (PCA) has been employed to identify independence and redundancies in multi-frequency eddy current data. This has allowed for an enhanced visualization of factors affecting gap measurement. Results of the PCA of simulation data are consistent with the skin depth equation, and are validated against PCA of physical experiments. Finally, compressed data acquisition has been realized, allowing faster data acquisition for multi-frequency eddy current systems with hardware limitations, and is generalizable to other applications where real time acquisition of large data sets is prohibitive.

Prise en compte d'un couplage fin neutronique-thermique dans les calculs d'assemblage pour les reacteurs a eau pressurisee

NASA Astrophysics Data System (ADS)

Greiner, Nathan

Core simulations for Pressurized Water Reactors (PWR) is insured by a set of computer codes which allows, under certain assumptions, to approximate the physical quantities of interest, such as the effective multiplication factor or the power or temperature distributions. The neutronics calculation scheme relies on three great steps : -- the production of an isotopic cross-sections library ; -- the production of a reactor database through the lattice calculation ; -- the full-core calculation. In the lattice calculation, in which Boltzmann's transport equation is solved over an assembly geometry, the temperature distribution is uniform and constant during irradiation. This represents a set of approximations since, on the one hand, the temperature distribution in the assembly is not uniform (strong temperature gradients in the fuel pins, discrepancies between the fuel pins) and on the other hand, irradiation causes the thermal properties of the pins to change, which modifies the temperature distribution. Our work aims at implementing and introducing a neutronics-thermomechanics coupling into the lattice calculation to finely discretize the temperature distribution and to study its effects. To perform the study, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) lattice code APOLLO2 was used for neutronics and EDF (Electricite De France) code C3THER was used for the thermal calculations. We show very small effects of the pin-scaled coupling when comparing the use of a temperature profile with the use of an uniform temperature over UOX-type and MOX-type fuels. We next investigate the thermal feedback using an assembly-scaled coupling taking into account the presence of large water gaps on an UOX-type assembly at burnup 0. We show the very small impact on the calculation of the hot spot factor. Finally, the coupling is introduced into the isotopic depletion calculation and we show that reactivity and isotopic number densities deviations remain small albeit not negligible for UOX-type and MOX-type assemblies. The specific behavior of gadolinium-stuffed fuel pins in an UO2Gd2O 3-type assembly is highlighted.

Developpement de mesures non destructives, par ondes ultrasonores, d'epaisseurs de fronts de solidification dans les reacteurs metallurgiques

NASA Astrophysics Data System (ADS)

Floquet, Jimmy

Dans les cuves d'electrolyse d'aluminium, le milieu de reaction tres corrosif attaque les parois de la cuve, ce qui diminue leur duree de vie et augmente les couts de production. Le talus, qui se forme sous l'effet des pertes de chaleur qui maintiennent un equilibre thermique dans la cuve, sert de protection naturelle a la cuve. Son epaisseur doit etre controlee pour maximiser cet effet. Advenant la resorption non voulue de ce talus, les degats generes peuvent s'evaluer a plusieurs centaines de milliers de dollars par cuve. Aussi, l'objectif est de developper une mesure ultrasonore de l'epaisseur du talus, car elle serait non intrusive et non destructive. La precision attendue est de l'ordre du centimetre pour des mesures d'epaisseurs comprenant 2 materiaux, allant de 5 a 20 cm. Cette precision est le facteur cle permettant aux industriels de controler l'epaisseur du talus de maniere efficace (maximiser la protection des parois tout en maximisant l'efficacite energetique du procede), par l'ajout d'un flux thermique. Cependant, l'efficacite d'une mesure ultrasonore dans cet environnement hostile reste a demontrer. Les travaux preliminaires ont permis de selectionner un transducteur ultrasonore a contact ayant la capacite a resister aux conditions de mesure (hautes temperatures, materiaux non caracterises...). Differentes mesures a froid (traite par analyse temps-frequence) ont permis d'evaluer la vitesse de propagation des ondes dans le materiau de la cuve en graphite et de la cryolite, demontrant la possibilite d'extraire l'information pertinente d'epaisseur du talus in fine. Fort de cette phase de caracterisation des materiaux sur la reponse acoustique des materiaux, les travaux a venir ont ete realises sur un modele reduit de la cuve. Le montage experimental, un four evoluant a 1050 °C, instrumente d'une multitude de capteurs thermique, permettra une comparaison de la mesure intrusive LVDT a celle du transducteur, dans des conditions proches de la mesure industrielle. Mots-cles : Ultrasons, CND, Haute temperature, Aluminium, Cuve d'electrolyse.

Simulation des fuites neutroniques a l'aide d'un modele B1 heterogene pour des reacteurs a neutrons rapides et a eau legere

NASA Astrophysics Data System (ADS)

Faure, Bastien

The neutronic calculation of a reactor's core is usually done in two steps. After solving the neutron transport equation over an elementary domain of the core, a set of parameters, namely macroscopic cross sections and potentially diffusion coefficients, are defined in order to perform a full core calculation. In the first step, the cell or assembly is calculated using the "fundamental mode theory", the pattern being inserted in an infinite lattice of periodic structures. This simple representation allows a precise modeling for the geometry and the energy variable and can be treated within transport theory with minimalist approximations. However, it supposes that the reactor's core can be treated as a periodic lattice of elementary domains, which is already a big hypothesis, and cannot, at first sight, take into account neutron leakage between two different zones and out of the core. The leakage models propose to correct the transport equation with an additional leakage term in order to represent this phenomenon. For historical reasons, numerical methods for solving the transport equation being limited by computer's features (processor speeds and memory sizes), the leakage term is, in most cases, modeled by a homogeneous and isotropic probability within a "homogeneous leakage model". Driven by technological innovation in the computer science field, "heterogeneous leakage models" have been developed and implemented in several neutron transport calculation codes. This work focuses on a study of some of those models, including the TIBERE model from the DRAGON-3 code developed at Ecole Polytechnique de Montreal, as well as the heterogeneous model from the APOLLO-3 code developed at Commissariat a l'Energie Atomique et aux energies alternatives. The research based on sodium cooled fast reactors and light water reactors has allowed us to demonstrate the interest of those models compared to a homogeneous leakage model. In particular, it has been shown that a heterogeneous model has a significant impact on the calculation of the out of core leakage rate that permits a better estimation of the transport equation eigenvalue Keff . The neutron streaming between two zones of different compositions was also proven to be better calculated.

A beta-ray spectrometer based on a two-or three silicon detector coincidence telescope

NASA Astrophysics Data System (ADS)

Horowitz, Y. S.; Weizman, Y.; Hirning, C. R.

1996-02-01

This report describes the operation of a beta-ray energy spectrometer based on a silicon detector telescope using two or three elements. The front detector is a planar, totally-depleted, silicon surface barrier detector that is 97 μm thick, the back detector is a room-temperature, lithium compensated, silicon detector that is 5000 μm thick, and the intermediate detector is similar to the front detector but 72 μm thick and intended to be used only in intense photon fields. The three detectors are mounted in a light-tight aluminum housing. The capability of the spectrometer to reject photons is based upon the fact that the incident photon will have a small probability of simultaneously losing detectable energy in two detectors, and an even smaller probability of losing detectable energy in all three detectors. Electrons will, however, almost always record measurable events in either the front two or all three detectors. A coincidence requirement between the detectors thus rejects photon induced events. With a 97 μm thick detector the lower energy coincidence threshold is approximately 110 keV. With an ultra-thin 40 μm thick front detector, and operated at 15°C, the spectrometer is capable of detecting even 60-70 keV electrons with a coincidence efficiency of 60%. The spectrometer has been used to measure beta radiation fields in CANDU reactor working environments, and the spectral information is intended to support dose algorithms for the LiF TLD chips used in the Ontario Hydro dosimetry program.

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

Techniques d'inspection par ondes guidees ultrasonores d'assemblages brases dans des reacteurs aeronautiques =

NASA Astrophysics Data System (ADS)

Comot, Pierre

L'industrie aeronautique, cherche a etudier la possibilite d'utiliser de maniere structurelle des joints brases, dans une optique de reduction de poids et de cout. Le developpement d'une methode d'evaluation rapide, fiable et peu couteuse pour evaluer l'integrite structurelle des joints apparait donc indispensable. La resistance mecanique d'un joint brase dependant principalement de la quantite de phase fragile dans sa microstructure. Les ondes guidees ultrasonores permettent de detecter ce type de phase lorsqu'elles sont couplees a une mesure spatio-temporelle. De plus la nature de ce type d'ondes permet l'inspection de joints ayant des formes complexes. Ce memoire se concentre donc sur le developpement d'une technique basee sur l'utilisation d'ondes guidees ultrasonores pour l'inspection de joints brases a recouvrement d'Inconel 625 avec comme metal d'apport du BNi-2. Dans un premiers temps un modele elements finis du joint a ete utilise pour simuler la propagation des ultrasons et optimiser les parametres d'inspection, la simulation a permis egalement de demontrer la faisabilite de la technique pour la detection de la quantite de phase fragile dans ce type de joints. Les parametres optimises sont la forme de signal d'excitation, sa frequence centrale et la direction d'excitation. Les simulations ont montre que l'energie de l'onde ultrasonore transmise a travers le joint aussi bien que celle reflechie, toutes deux extraites des courbes de dispersion, etaient proportionnelles a la quantite de phase fragile presente dans le joint et donc cette methode permet d'identifier la presence ou non d'une phase fragile dans ce type de joint. Ensuite des experimentations ont ete menees sur trois echantillons typiques presentant differentes quantites de phase fragile dans le joint, pour obtenir ce type d'echantillons differents temps de brasage ont ete utilises (1, 60 et 180 min). Pour cela un banc d'essai automatise a ete developpe permettant d'effectuer une analyse similaire a celle utilisee en simulation. Les parametres experimentaux ayant ete choisis en accord avec l'optimisation effectuee lors des simulations et apres une premiere optimisation du procede experimental. Finalement les resultats experimentaux confirment les resultats obtenus en simulation, et demontrent le potentiel de la methode developpee.

The General Formulation and Practical Calculation of the Diffusion Coefficient in a Lattice Containing Cavities; FORMULATION GENERALE ET CALCUL PRATIQUE DU COEFFICIENT DE DIFFUSION DANS UN RESEAU COMPORTANT DES CAVITES (in French)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoist, P.

The calculation of diffusion coefficients in a lattice necessitates the knowledge of a correct method of weighting the free paths of the different constituents. An unambiguous definition of this weighting method is given here, based on the calculation of leakages from a zone of a reactor. The formulation obtained, which is both simple and general, reduces the calculation of diffusion coefficients to that of collision probabilities in the different media; it reveals in the expression for the radial coefficient the series of the terms of angular correlation (cross terms) recently shown by several authors. This formulation is then used tomore » calculate the practical case of a classical type of lattice composed of a moderator and a fuel element surrounded by an empty space. Analytical and numerical comparison of the expressions obtained with those inferred from the theory of BEHRENS shows up the importance of several new terms some of which are linked with the transparency of the fuel element. Cross terms up to the second order are evaluated. A practical formulary is given at the end of the paper. (author) [French] Le calcul des coefficients de diffusion dans un reseau suppose la connaissance d'un mode de ponderation correct des libres parcours des differents constituants. On definit ici sans ambiguite ce mode de ponderation a partir du calcul des fuites hors d'une zone de reacteur. La formulation obtenue, simple et generale, ramene le calcul des coefficients de diffusion a celui des probabilites de collision dans les differents milieux; elle fait apparaitre dans l'expression du coefficient radial la serie des termes de correlation angulaire (termes rectangles), mis en evidence recemment par plusieurs auteurs. Cette formulation est ensuite appliquee au calcul pratique d'un reseau classique, compose d'un moderateur et d'un element combustible entoure d'une cavite; la comparaison analytique et numerique des expressions obtenues avec celles deduites de la theorie de BEHRENS fait apparaitre l'importance de plusieurs termes nouveaux, dont certains sont lies a la transparence de l'element combustible; les termes rectangles sont calcules jusqu'a l'ordre 2. Un formulaire pratique est donne a la fin de cette etude. (auteur)« less

Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

NASA Astrophysics Data System (ADS)

Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

2018-04-01

Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

Knowledge elicitation techniques and application to nuclear plant maintenance

NASA Astrophysics Data System (ADS)

Doyle, E. Kevin

The new millennium has brought with it the opportunity of global trade which in turn requires the utmost in efficiency from each individual industry. This includes the nuclear power industry, a point which was emphasized when the electrical generation industry began to be de regulated across North America the late 1990s and re-emphasized when the northeast power grid of North America collapsed in the summer of 2003. This dissertation deals with reducing the cost of the maintenance function of Candu nuclear power plants and initiating a strong link between universities and the Canadian nuclear industry. Various forms of RCM (reliability-centred maintenance) have been the tools of choice in industry for improving the maintenance function during the last 20 years. In this project, pilot studies, conducted at Bruce Power between 1999 and 2005, and reported on in this dissertation, lay out a path to implement statistical improvements as the next step after RCM in reducing the cost of the maintenance. Elicitation protocols, designed for the age group being elicited, address the much-documented issue of a lack of data. Clear, graphical, inferential statistical interfaces are accentuated and developed to aid in building the teams required to implement the various methodologies and to help in achieving funding targets. Graphical analysis and Crow/AMSAA (army materials systems analysis activity) plots are developed and demonstrated from the point of view of justifying the expenditures of cost reduction efforts. This dissertation ultimately speaks to the great opportunity being presented by this approach at this time: of capturing the baby-boom generation's huge pool of knowledge before those people retire. It is expected that the protocols and procedures referenced here will have applicability across the many disciplines where collecting expert information from a similar age group is required.

Remote Field Eddy Curent Signal Modeling for the Gap Measurement of Neighboring Tubes

NASA Astrophysics Data System (ADS)

Jung, H. K.; Lee, D. H.; Lee, Y. S.

2005-04-01

The fuel channels in the Canadian Deuterium Uranium (CANDU) reactor consist of the coaxial pressure tube (PT) and the calandria tube (CT). The Liquid injection nozzle (LIN) is cross aligned with the fuel channel to control the reactor by injecting poison. For a safe operation, the gap between the LIN and CT should be maintained in order to prevent a contact of the neighboring tubes. The remote field eddy current (RFEC) method was applied to measure the gap between a nonmagnetic Zircaloy-2 liquid injection nozzle (LIN) and a Zircaloy-2 calandria tube. Under the condition of inserting the RFEC probe into the coaxial tubes and of crossing a LIN above or under the CT, the modeling of a LIN signal is needed to check the possibility of a gap measurement. The Volume Integral Code S/W which covers the axi-symmetric 3D configuration has been very rarely applied to obtain a LIN signal. This problem was solved by assuming a LIN as a flaw which can be described as a complete 3D object. This simulated LIN signal was verified by performing the laboratory experiment. The gap between the LIN and CT can be correlated with the amplitude of the LIN signals in the voltage plane. Typical noises in the fuel channel were the relative constriction, the change in the pressure tube diameter (fill-factor), thickness variation, and so on. These noise signals were simulated by using the modeling and were analyzed by considering their dependency on the phase angle and amplitude of the voltage plane in order to separate the gap signal from them. It could be concluded that the voltage plane analysis of the simulated RFEC signals were effective for obtaining the gap measurement of the neighboring tube.

Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

PubMed

Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

2013-09-01

Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

Application of perturbation theory to lattice calculations based on method of cyclic characteristics

NASA Astrophysics Data System (ADS)

Assawaroongruengchot, Monchai

Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended from 300 to 450 FPDs. The cases with the central pin containing either Dysprosium or Gadolinium in the natural Uranium are considered in our study. (Abstract shortened by UMI.)

Comparison of Calibration of Sensors Used for the Quantification of Nuclear Energy Rate Deposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brun, J.; Reynard-Carette, C.; Tarchalski, M.

This present work deals with a collaborative program called GAMMA-MAJOR 'Development and qualification of a deterministic scheme for the evaluation of GAMMA heating in MTR reactors with exploitation as example MARIA reactor and Jules Horowitz Reactor' between the National Centre for Nuclear Research of Poland, the French Atomic Energy and Alternative Energies Commission and Aix Marseille University. One of main objectives of this program is to optimize the nuclear heating quantification thanks to calculation validated from experimental measurements of radiation energy deposition carried out in irradiation reactors. The quantification of the nuclear heating is a key data especially for themore » thermal, mechanical design and sizing of irradiation experimental devices in specific irradiated conditions and locations. The determination of this data is usually performed by differential calorimeters and gamma thermometers such as used in the experimental multi-sensors device called CARMEN 'Calorimetric en Reacteur et Mesures des Emissions Nucleaires'. In the framework of the GAMMA-MAJOR program a new calorimeter was designed for the nuclear energy deposition quantification. It corresponds to a single-cell calorimeter and it is called KAROLINA. This calorimeter was recently tested during an irradiation campaign inside MARIA reactor in Poland. This new single-cell calorimeter differs from previous CALMOS or CARMEN type differential calorimeters according to three main points: its geometry, its preliminary out-of-pile calibration, and its in-pile measurement method. The differential calorimeter, which is made of two identical cells containing heaters, has a calibration method based on the use of steady thermal states reached by simulating the nuclear energy deposition into the calorimeter sample by Joule effect; whereas the single-cell calorimeter, which has no heater, is calibrated by using the transient thermal response of the sensor (heating and cooling steps). The paper will concern these two kinds of calorimetric sensors. It will focus in particular on studies on their out-of-pile calibrations. Firstly, the characteristics of the sensor designs will be detailed (such as geometry, dimension, material sample, assembly, instrumentation). Then the out-of-pile calibration methods will be described. Furthermore numerical results obtained thanks to 2D axisymmetrical thermal simulations (Finite Element Method, CAST3M) and experimental results will be presented for each sensor. A comparison of the two different thermal sensor behaviours will be realized. To conclude a discussion of the advantages and the drawbacks of each sensor will be performed especially regarding measurement methods. (authors)« less

Restructuring the Uranium Mining Industry in Romania: Actual Situation and Prospects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Georgescu, P.D.; Petrescu, S.T.; Iuhas, T.F.

2002-07-01

Uranium prospecting in Romania has started some 50 years ago, when a bilateral agreement between Romania and the former Soviet Union had been concluded and a joint Romanian-Soviet enterprise was created. The production started in 1952 by the opening of some deposits from western Transylvania (Bihor and Ciudanovita). From 1962 the production has continued only with Romanian participation on the ore deposit Avram Iancu and from 1985 on the deposits from Eastern Carpathians (Crucea and Botusana). Starting with 1978 the extracted ores have been completely processed in the Uranium Ore Processing Plant from Feldioara, Brasov. Complying with the initial stipulationsmore » of the Nuclear National Program launched at the beginning of the 1980's, the construction of a nuclear power station in Cernavoda has started in Romania, using natural uranium and heavy water (CANDU type), having five units of 650 MW installed capacity. After 1989 this initial Nuclear National Program was revised and the construction of the first unit (number 1) was finalized and put in operation in 1996. In 2001 the works at the unit number 2 were resumed, having the year 2005 as the scheduled activating date. The future of the other 3 units, being in different construction phases, hasn't been clearly decided. Taking into consideration the exhaustion degree of some ore deposits and from the prospect of exploiting other ore deposits, the uranium industry will be subject of an ample restructuring process. This process includes workings of modernization of the mines in operation and of the processing plant, increasing the profitableness, lowering of the production costs by closing out and ecological rehabilitation of some areas affected by mining works and even new openings of some uraniferous exploitations. This paper presents the actual situation and the prospects of uranium mining industry on the base of some new technical and economical strategic concepts in accordance with the actual Romanian Program for Nuclear Energetics. (authors)« less

Investigation of Dispersed and Dispersed Annular (rivulet or Thin Film) Flow Phase Separation in Tees.

NASA Astrophysics Data System (ADS)

McCreery, Glenn Ernest

An experimental and analytical investigation of dispersed and dispersed-annular (rivulet or thin film) flow phase separation in tees has been successfully completed. The research was directed at, but is not specific to, determining flow conditions, following a loss of coolant accident, in the large rectangular passageways leading to vacuum buildings in the containment envelope of some CANDU nuclear reactors. The primary objectives of the research were to: (1) obtain experimental data to help formulate and test mechanistic analytical models of phase separation, and (2) develop the analytical models in computer programs which predict phase separation from upstream flow and pressure conditions and downstream and side branch pressure boundary conditions. To meet these objectives an air-water experimental apparatus was constructed, and consists of large air blowers attached to a long rectangular duct leading to a tee in the horizontal plane. A variety of phenomena was investigated including, for comparison with computer predictions, air streamlines and eddy boundary geometry, drop size spectra, macroscopic mass balances, liquid rivulet pathlines, and trajectories of drops of known size and velocity. Four separate computer programs were developed to analyze phase separation. Three of the programs are used sequentially to calculate dispersed mist phase separation in a tee. The fourth is used to calculate rivulet or thin film pathlines. Macroscopic mass balances are calculated from a summation of mass balances for drops with representative sizes (and masses) spaced across the drop size spectrum. The programs are tested against experimental data, and accurately predict gas flow fields, drop trajectories, rivulet pathlines and macroscopic mass balances. In addition to development of the computer programs, analysis was performed to specify the scaling of dispersed mist and rivulet or thin film flow, to investigate pressure losses in tees, and the inter-relationship of loss coefficients, contraction coefficients, and eddy geometry. The important transient effects of liquid storage in eddies were also analyzed.

Microstructural evolution of CANDU spacer material Inconel X-750 under in situ ion irradiation

NASA Astrophysics Data System (ADS)

Zhang, He Ken; Yao, Zhongwen; Judge, Colin; Griffiths, Malcolm

2013-11-01

Work on Inconel®Inconel® is a registered trademark of Special Metals Corporation that refers to a family of austenitic nickel-chromium-based superalloys.1 X-750 spacers removed from CANDU®CANDU® is a registered trademark of Atomic Energy of Canada Limited standing for ''CANada Deuterium Uranium''.2 reactors has shown that they become embrittled and there is development of many small cavities within the metal matrix and along grain boundaries. In order to emulate the neutron irradiation induced microstructural changes, heavy ion irradiations (1 MeV Kr2+ ions) were performed while observing the damage evolution using an intermediate voltage electron microscope (IVEM) operating at 200 kV. The irradiations were carried out at various temperatures 60-400 °C. The principal strengthening phase, γ‧, was disordered at low doses (˜0.06 dpa) during the irradiation. M23C6 carbides were found to be stable up to 5.4 dpa. Lattice defects consisted mostly of stacking fault tetrahedras (SFTs), 1/2<1 1 0> perfect loops and small 1/3<1 1 1> faulted Frank loops. The ratio of SFT number density to loop number density for each irradiation condition was found to be neither temperature nor dose dependent. Under the operation of the ion beam the SFT production was very rapid, with no evidence for further growth once formed, indicating that they probably formed as a result of cascade collapse in a single cascade. The number density of the defects was found to saturate at low dose (˜0.68 dpa). No cavities were observed regardless of the irradiation temperature between 60 °C and 400 °C for doses up to 5.4 dpa. In contrast, cavities have been observed after neutron irradiation in the same material at similar doses and temperatures indicating that helium, produce during neutron irradiation, may be essential for the nucleation and growth of cavities.

Validation d'un nouveau calcul de reference en evolution pour les reacteurs thermiques

NASA Astrophysics Data System (ADS)

Canbakan, Axel

Resonance self-shielding calculations are an essential component of a deterministic lattice code calculation. Even if their aim is to correct the cross sections deviation, they introduce a non negligible error in evaluated parameters such as the flux. Until now, French studies for light water reactors are based on effective reaction rates obtained using an equivalence in dilution technique. With the increase of computing capacities, this method starts to show its limits in precision and can be replaced by a subgroup method. Originally used for fast neutron reactor calculations, the subgroup method has many advantages such as using an exact slowing down equation. The aim of this thesis is to suggest a validation as precise as possible without burnup, and then with an isotopic depletion study for the subgroup method. In the end, users interested in implementing a subgroup method in their scheme for Pressurized Water Reactors can rely on this thesis to justify their modelization choices. Moreover, other parameters are validated to suggest a new reference scheme for fast execution and precise results. These new techniques are implemented in the French lattice scheme SHEM-MOC, composed of a Method Of Characteristics flux calculation and a SHEM-like 281-energy group mesh. First, the libraries processed by the CEA are compared. Then, this thesis suggests the most suitable energetic discretization for a subgroup method. Finally, other techniques such as the representation of the anisotropy of the scattering sources and the spatial representation of the source in the MOC calculation are studied. A DRAGON5 scheme is also validated as it shows interesting elements: the DRAGON5 subgroup method is run with a 295-eenergy group mesh (compared to 361 groups for APOLLO2). There are two reasons to use this code. The first involves offering a new reference lattice scheme for Pressurized Water Reactors to DRAGON5 users. The second is to study parameters that are not available in APOLLO2 such as self-shielding in a temperature gradient and using a flux calculation based on MOC in the self-shielding part of the simulation. This thesis concludes that: (1) The subgroup method is at least more precise than a technique based on effective reaction rates, only if we use a 361-energy group mesh; (2) MOC with a linear source in a geometrical region gives better results than a MOC with a constant model. A moderator discretization is compulsory; (3) A P3 choc law is satisfactory, ensuring a coherence with 2D full core calculations; (4) SHEM295 is viable with a Subgroup Projection Method for DRAGON5.

Evolution of interphase and intergranular strain in zirconium-niobium alloys during deformation at room temperature

NASA Astrophysics Data System (ADS)

Cai, Song

Zr-2.5Nb is currently used for pressure tubes in the CANDU (CANada Deuterium Uranium) reactor. A complete understanding of the deformation mechanism of Zr-2.5Nb is important if we are to accurately predict the in-reactor performance of pressure tubes and guarantee normal operation of the reactors. This thesis is a first step in gaining such an understanding; the deformation mechanism of ZrNb alloys at room temperature has been evaluated through studying the effect of texture and microstructure on deformation. In-situ neutron diffraction was used to monitor the evolution of the lattice strain of individual grain families along both the loading and Poisson's directions and to track the development of interphase and intergranular strains during deformation. The following experiments were carried out with data interpreted using elasto-plastic modeling techniques: (1) Compression tests of a 100%betaZr material at room temperature. (2) Tension and compression tests of hot rolled Zr-2.5Nb plate material. (3) Compression of annealed Zr-2.5Nb. (4) Cyclic loading of the hot rolled Zr-2.5Nb. (5) Compression tests of ZrNb alloys with different Nb and oxygen contents. The experimental results were interpreted using a combination of finite element (FE) and elasto-plastic self-consistent (EPSC) models. The phase properties and phase interactions well represented by the FE model, the EPSC model successfully captured the evolution of intergranular constraint during deformation and provided reasonable estimates of the critical resolved shear stress and hardening parameters of different slip systems under different conditions. The consistency of the material parameters obtained by the EPSC model allows the deformation mechanism at room temperature and the effect of textures and microstructures of ZrNb alloys to be understood. This work provides useful information towards manufacturing of Zr-2.5Nb components and helps in producing ideal microstructures and material properties for pressure tubes. Also it is helpful in guiding the development of new materials for the next generation of nuclear reactors. Furthermore, the large data set obtained from this study can be used in evaluation and improving current and future polycrystalline deformation models.

CFD Application to Flow-Accelerated Corrosion in Feeder Bends

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pietralik, John M.; Smith, Bruce A.W.

2006-07-01

Feeder piping in CANDU{sup R} plants experiences a thinning degradation mechanism called Flow-Accelerated Corrosion (FAC). The piping is made of carbon steel and has high water flow speeds. Although the water chemistry is highly alkaline with room-temperature pH in a range of 10.0-10.5, the piping has FAC rates exceeding 0.1 mm/year in some locations, e.g., in bends. One of the most important parameters affecting the FAC rate is the mass transfer coefficient for convective mass transport of ferrous ions. The ions are created at the pipe wall as a result of corrosion, diffuse through the oxide layer, and are transportedmore » from the oxide-layer/water interface to the bulk water by mass transport. Consequently, the local flow characteristics contribute to the highly turbulent convective mass transfer. Plant data and laboratory experiments indicate that the mass transfer step dominates FAC under feeder conditions. In this study, the flow and mass transfer in a feeder bend under operating conditions were simulated using the Fluent{sup TM} computer code. Because the flow speed is very high, with the Reynolds numbers in a range of several millions, and because the geometry is complex, experiments in a 1:1 scale were conducted with the main objective to validate flow simulations. The experiments measured pressure at several key locations and visualized the flow. The flow and mass transfer models were validated using available friction-factor and mass transfer correlations and literature experiments on mass transfer in a bend. The validation showed that the turbulence model that best predicts the experiments is the realizable k-{epsilon} model. Other two-equation turbulence models, as well as one-equation models and Reynolds stress models were tried. The near-wall treatment used the non-equilibrium wall functions. The wall functions were modified for surface roughness when necessary. A comparison of the local mass transfer coefficient with measured FAC rate in plant specimens shows very good agreement. Visualization experiments indicate secondary flows in the bends. No boundary layer separation was observed in experiments or in simulations. (authors)« less

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

Radiochemistry Lab Decommissioning and Dismantlement. AECL, Chalk River Labs, Ontario, Canada

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kenny, Stephen

2008-01-15

Atomic Energy of Canada (AECL) was originally founded in the mid 1940's to perform research in radiation and nuclear areas under the Canadian Defense Department. In the mid 50's The Canadian government embarked on several research and development programs for the development of the Candu Reactor. AECL was initially built as a temporary site and is now faced with many redundant buildings. Prior to 2004 small amounts of Decommissioning work was in progress. Many reasons for deferring decommissioning activities were used with the predominant ones being: 1. Reduction in radiation doses to workers during the final dismantlement, 2. Development ofmore » a long-term solution for the management of radioactive wastes in Canada, 3. Financial constraints presented by the number of facilities shutdown that would require decommissioning funds and the absence of an approved funding strategy. This has led to the development of a comprehensive decommissioning plan that is all inclusive of AECL's current and legacy liabilities. Canada does not have a long-term disposal site; therefore waste minimization becomes the driving factor behind decontamination for decommissioning before and during dismantlement. This decommissioning job was a great learning experience for decommissioning and the associated contractors who worked on this project. Throughout the life of the project there was a constant focus on waste minimization. This focus was constantly in conflict with regulatory compliance primarily with respect to fire regulations and protecting the facility along with adjacent facilities during the decommissioning activities. Discrepancies in historical documents forced the project to treat every space as a contaminated space until proven differently. Decommissioning and dismantlement within an operating site adds to the complexity of the tasks especially when it is being conducted in the heart of the plant. This project was very successful with no lost time accidents in over one hundred thousand hours worked, on schedule and under budget despite some significant changes throughout the decommissioning phases. The actual cost to decommission this building will come in under 9 million dollars vs. an estimated 14.5 million dollars. This paper will cover some of the unique aspects of dismantling a radioactive building that has seen pretty much every element of the periodic table pass through it with the client requirement focused on minimization of radioactive waste volumes.« less

Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Lizon-A-Lugrin, Laure

The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a mechanical power of 1200 MW. It is observed that in most cases the landscape of Pareto's front is mostly controlled only by few key parameters. These results may be very useful for future plant design engineers. Furthermore, some calculations for pipe sizing and temperature variation between coolant and fuel have been carried out to provide an idea on their order of magnitude.

Developpement d'un catalyseur nickel-alumine efficace pour le reformage de diesel a la vapeur d'eau et etude du systeme reactionnel

NASA Astrophysics Data System (ADS)

Fauteux-Lefebvre, Clemence

Le developpement de sources d'energie alternatives fiables et efficaces est aujourd'hui une necessite. L' interet dans le reformage d'hydrocarbures liquides est ainsi croissant puisqu'il s'agit d'une voie pour l'alimentation des piles a combustible. Les piles a combustible ont une efficacite pour la conversion d'energie en electricite plus grande que celle des moteurs a combustion et font ainsi partie de la recherche de solution en efficacite energetique. Ces piles consomment de l'hydrogene comme combustible pour produire de l'electricite, d'ou l'interet pour le reformage. En effet, cette reaction permet de produire de l'hydrogene et du monoxyde de carbone (un autre combustible des piles a combustible a electrolyte solide) a partir d'hydrocarbure liquide, notamment le diesel. Les piles pourraient donc etre integrees avec une unite de reformage leur fournissant directement le combustible necessaire a partir de diesel. Dans ce projet de recherche, un nouveau catalyseur de nickel sous forme de spinelle nickel-alumine (spinelle NiAl2O4 sur support d'alumine et de zircone stabilisee avec yttria) a ete developpe et teste en laboratoire pour du reformage de propane, d'hydrocarbures liquides et de diesel, a la vapeur d'eau. Par ailleurs, une methode d'ajout des reactifs novatrice a ete utilisee afin de diminuer la pyrolyse precedant le reformage, en utilisant une emulsion. Les resultats de reformage d'hydrocarbures purs ont montre des concentrations tres pres de l'equilibre thermodynamique et une activite constante sans desactivation du catalyseur ni formation de carbone, et ce avec des ratios H2O/C de moins de 2.5 et des temperatures d'operation variant entre 630 °C et 750 °C. Lors de tests effectues en utilisant du diesel fossile, a 705°C, avec un debit volumique des reactifs de plus de 50 000 cm3gcat-1h-1 et un ratio H2O/C de moins de 2.5, l'activite a ete maintenue pendant plus de 15 heures, malgre une operation en cycles. L'analyse du catalyseur apres cette utilisation n'a montre aucun carbone significatif sur la surface.- En comparaison, un catalyseur de nickel metallique sur support d'Al2O3 et YSZ a ete utilise dans des conditions similaires. Il y a eu desactivation du catalyseur et obstruction du reacteur par du carbone apres trois heures d'operation. L'analyse de ce catalyseur a permis de verifier qu'il etait recouvert de carbone en filament. L'analyse du systeme reactionnel a montre que la reaction est controlee par la reaction de. surface et non par le transfert de masse. Par ailleurs, les analyses des catalyseurs de spinelle ont demontre qu'il n'y avait pas de modification de sa forme chimique ni de reduction du spinelle en nickel metallique apres l'utilisation. Mots cles : Reformage a la vapeur, diesel, hydrocarbure liquide, catalyseur, spinelle nickel-alumine, equilibre thermodynamique

Risk-informed Maintenance for Non-coherent Systems

NASA Astrophysics Data System (ADS)

Tao, Ye

Probabilistic Safety Assessment (PSA) is a systematic and comprehensive methodology to evaluate risks associated with a complex engineered technological entity. The information provided by PSA has been increasingly implemented for regulatory purposes but rarely used in providing information for operation and maintenance activities. As one of the key parts in PSA, Fault Tree Analysis (FTA) attempts to model and analyze failure processes of engineering and biological systems. The fault trees are composed of logic diagrams that display the state of the system and are constructed using graphical design techniques. Risk Importance Measures (RIMs) are information that can be obtained from both qualitative and quantitative aspects of FTA. Components within a system can be ranked with respect to each specific criterion defined by each RIM. Through a RIM, a ranking of the components or basic events can be obtained and provide valuable information for risk-informed decision making. Various RIMs have been applied in various applications. In order to provide a thorough understanding of RIMs and interpret the results, they are categorized with respect to risk significance (RS) and safety significance (SS) in this thesis. This has also tied them into different maintenance activities. When RIMs are used for maintenance purposes, it is called risk-informed maintenance. On the other hand, the majority of work produced on the FTA method has been concentrated on failure logic diagrams restricted to the direct or implied use of AND and OR operators. Such systems are considered as coherent systems. However, the NOT logic can also contribute to the information produced by PSA. The importance analysis of non-coherent systems is rather limited, even though the field has received more and more attention over the years. The non-coherent systems introduce difficulties in both qualitative and quantitative assessment of the fault tree compared with the coherent systems. In this thesis, a set of RIMs is analyzed and investigated. The 8 commonly used RIMs (Birnbaum's Measure, Criticality Importance Factor, Fussell-Vesely Measure, Improvement Potential, Conditional Probability, Risk Achievement, Risk Achievement Worth, and Risk Reduction Worth) are extended to non-coherent forms. Both coherent and non-coherent forms are classified into different categories in order to assist different types of maintenance activities. The real systems such as the Steam Generator Level Control System in CANDU Nuclear Power Plant (NPP), a Gas Detection System, and the Automatic Power Control System of the experimental nuclear reactor are presented to demonstrate the application of the results as case studies.

Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion irradiated hot-rolled Zr-2.5Nb alloy is described, providing evidence for the interaction between moving dislocations and irradiation induced loops. Chapter 8 gives the effect on the dislocation structure of different levels of compressive strains along two directions in the hot-rolled Zr-2.5Nb alloy. By using high resolution neutron diffraction and TEM observations, the evolution of type and dislocation densities, as well as changes of dislocation microstructure with plastic strain were characterized.

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Radiochemical Separation and Quantification of Tritium in Metallic Radwastes Generated from CANDU Type NPP - 13279

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, H.J.; Choi, K.C.; Choi, K.S.

2013-07-01

As a destructive quantification method of {sup 3}H in low and intermediate level radwastes, bomb oxidation, sample oxidation, and wet oxidation methods have been introduced. These methods have some merits and demerits in the radiochemical separation of {sup 3}H radionuclides. That is, since the bomb oxidation and sample oxidation methods are techniques using heating at high temperature, the separation methods of the radionuclides are relatively simple. However, since {sup 3}H radionuclide has a property of being diffused deeply into the inside of metals, {sup 3}H which is distributed on the surface of the metals can only be extracted if themore » methods are applied. As an another separation method, the wet oxidation method makes {sup 3}H oxidized with an acidic solution, and extracted completely to an oxidized HTO compound. However, incomplete oxidized {sup 3}H compounds, which are produced by reactions of acidic solutions and metallic radwastes, can be released into the air. Thus, in this study, a wet oxidation method to extract and quantify the {sup 3}H radionuclide from metallic radwastes was established. In particular, a complete extraction method and complete oxidation method of incomplete chemical compounds of {sup 3}H using a Pt catalyst were studied. The radioactivity of {sup 3}H in metallic radwastes is extracted and measured using a wet oxidation method and liquid scintillation counter. Considering the surface dose rate of the sample, the appropriate size of the sample was determined and weighed, and a mixture of oxidants was added to a 200 ml round flask with 3 tubes. The flask was quickly connected to the distilling apparatus. 20 mL of 16 wt% H{sub 2}SO{sub 4} was given into the 200-ml round flask through a dropping funnel while under stirring and refluxing. After dropping, the temperature of the mixture was raised to 96 deg. C and the sample was leached and oxidized by refluxing for 3 hours. At that time, the incomplete oxidized {sup 3}H compounds were completely oxidized using the Pt catalysts and produced a stable HTO compound. After that, about a 20 ml solution was distilled in the separation apparatus, and the distillate was mixed with an ultimagold LLT as a cocktail solution. The solution in the vial was left standing for at least 24 hours. The radioactivity of {sup 3}H was counted directly using a liquid scintillation analyzer (Packard, 2500 TR/AB, Alpha and Beta Liquid Scintillation Analyzer). (authors)« less

Effects of the Fabrication Process and Thermal Cycling on the Oxidation of Zirconium-Niobium Pressure Tubes

NASA Astrophysics Data System (ADS)

Nam, Cheol

2009-12-01

Pressure tubes made of Zr-2.5%Nb alloy are used to contain fuels and coolant in CANDU nuclear power reactors The pressure tube oxidizes during reactor operation and hydrogen ingress through the oxide grown on the tube limits its lifetime. Little attention was paid to the intermediate tube manufacturing processes in enhancing the oxidation resistance. In addition, the oxide grown on the tube experiences various thermal cycles depending on the reactor shutdown and startup cycles. To address these two aspects and to better understand the oxidation process of the Zr-2.5Nb tube, research was conducted in two parts: (i) effects of tube fabrication on oxidation behavior, and (ii) thermal cycling behaviors of oxides grown on a pressure tube. In the first part, the optimum manufacturing process was pursued to improve the corrosion resistance of Zr-2.5Nb tubes. Experimental micro-tubes were fabricated with various manufacturing routes in the stages of billet preparation, hot extrusion and cold drawing. These were oxidized in air at 400°C and 500°C, and in an autoclave at 360°C lithiated water. Microstructure and texture of the tubes and oxides were characterized with X-ray diffraction, scanning electron microscope and optical microscope. Special emphasis was given to examinations of the metal/oxide interface structures. A correlation between the manufacturing process and oxidation resistance was investigated in terms of tube microstructure and the metal/oxide interface structure. As a result, it was consistently observed that uniform interface structures were formed on the tubes which had a fine distribution of secondary phases. These microstructures were found to be beneficial in enhancing the oxidation resistance as opposed to the tubes that had coarse and continuous beta-Zr phases. Based on these observations, a schematic model of the oxidation process was proposed with respect to the oxidation resistance under oxidizing temperatures of 360°C, 400°C and 500°C. In the second part, the oxides grown on a standard Zr-2.5Nb pressure tube were analyzed by X-ray diffraction peak broadening and line shift. Crystallite size, t-ZrO2 fraction and residual stress of the zirconium oxides were investigated upon several thermal cycles at DeltaT range of 500°C--750°C. The oxide residual stresses measured by the sin2psi method were always compressive around 2 GPa. Different stress-states were noticed with the oxides grown on different sections of pressure tube. The compressive stress was released when the oxide was thermally cycled at the highest DeltaT of 750°C. Discussion was given to the effects of anisotropic nature of thermal expansion coefficients and crystallographic texture on the stress-state of Zr oxides.

Methodes d'optimisation des parametres 2D du reflecteur dans un reacteur a eau pressurisee

NASA Astrophysics Data System (ADS)

Clerc, Thomas

With a third of the reactors in activity, the Pressurized Water Reactor (PWR) is today the most used reactor design in the world. This technology equips all the 19 EDF power plants. PWRs fit into the category of thermal reactors, because it is mainly the thermal neutrons that contribute to the fission reaction. The pressurized light water is both used as the moderator of the reaction and as the coolant. The active part of the core is composed of uranium, slightly enriched in uranium 235. The reflector is a region surrounding the active core, and containing mostly water and stainless steel. The purpose of the reflector is to protect the vessel from radiations, and also to slow down the neutrons and reflect them into the core. Given that the neutrons participate to the reaction of fission, the study of their behavior within the core is capital to understand the general functioning of how the reactor works. The neutrons behavior is ruled by the transport equation, which is very complex to solve numerically, and requires very long calculation. This is the reason why the core codes that will be used in this study solve simplified equations to approach the neutrons behavior in the core, in an acceptable calculation time. In particular, we will focus our study on the diffusion equation and approximated transport equations, such as SPN or S N equations. The physical properties of the reflector are radically different from those of the fissile core, and this structural change causes important tilt in the neutron flux at the core/reflector interface. This is why it is very important to accurately design the reflector, in order to precisely recover the neutrons behavior over the whole core. Existing reflector calculation techniques are based on the Lefebvre-Lebigot method. This method is only valid if the energy continuum of the neutrons is discretized in two energy groups, and if the diffusion equation is used. The method leads to the calculation of a homogeneous reflector. The aim of this study is to create a computational scheme able to compute the parameters of heterogeneous, multi-group reflectors, with both diffusion and SPN/SN operators. For this purpose, two computational schemes are designed to perform such a reflector calculation. The strategy used in both schemes is to minimize the discrepancies between a power distribution computed with a core code and a reference distribution, which will be obtained with an APOLLO2 calculation based on the method Method Of Characteristics (MOC). In both computational schemes, the optimization parameters, also called control variables, are the diffusion coefficients in each zone of the reflector, for diffusion calculations, and the P-1 corrected macroscopic total cross-sections in each zone of the reflector, for SPN/SN calculations (or correction factors on these parameters). After a first validation of our computational schemes, the results are computed, always by optimizing the fast diffusion coefficient for each zone of the reflector. All the tools of the data assimilation have been used to reflect the different behavior of the solvers in the different parts of the core. Moreover, the reflector is refined in six separated zones, corresponding to the physical structure of the reflector. There will be then six control variables for the optimization algorithms. [special characters omitted]. Our computational schemes are then able to compute heterogeneous, 2-group or multi-group reflectors, using diffusion or SPN/SN operators. The optimization performed reduces the discrepancies distribution between the power computed with the core codes and the reference power. However, there are two main limitations to this study: first the homogeneous modeling of the reflector assemblies doesn't allow to properly describe its physical structure near the core/reflector interface. Moreover, the fissile assemblies are modeled in infinite medium, and this model reaches its limit at the core/reflector interface. These two problems should be tackled in future studies. (Abstract shortened by UMI.).

Modelisation de l'instabilite fluidelastique d'un faisceau de tubes soumis a un ecoulement diphasique transverse

NASA Astrophysics Data System (ADS)

Sawadogo, Teguewinde

This study focuses on the modeling of fluidelastic instability induced by two-phase cross-flow in tube bundles of steam generators. The steam generators in CANDU type nuclear power plants for e.g., designed in Canada by AECL and exploited worldwide, have thousands of tubes assembled in bundles that ensure the heat exchange between the internal circuit of heated heavy water coming from the reactor core and the external circuit of light water evaporated and directed toward the turbines. The main objective of this research project is to extend the theoretical models for fluidelastic instability to two-phase flow, validate the models and develop a computer program for simulating flow induced vibrations in tube bundles. The quasi-steady model has been investigated in scope of this research project. The time delay between the structure motion and the fluid forces generated thereby has been extensively studied in two-phase flow. The study was conducted for a rotated triangular tube array. Firstly, experimental measurements of unsteady and quasi-static fluid forces (in the lift direction) acting on a tube subject to two-phase flow were conducted. Quasi-static fluid force coefficients were measured at the same Reynolds number, Re = 2.8x104, for void fractions ranging from 0% to 80%. The derivative of the lift coefficient with respect to the quasi-static dimensionless displacement in the lift direction was deduced from the experimental measurements. This derivative is one of the most important parameters of the quasi-steady model because this parameter, in addition to the time delay, generates the fluid negative damping that causes the instability. This derivative was found to be positive in liquid flow and negative in two-phase flow. It seemed to vanish at 5% of void fraction, challenging the ability of the quasi-steady model to predict fluidelastic instability in this case. However, stability tests conducted at 5% void fraction clearly showed fluidelastic instability. Stability tests were conducted in the second stage of the project to validate the theoretical model. The two phase damping, the added mass and the critical velocity for fluidelastic instability were measured in two-phase flow. A viscoelastic damper was designed to vary the damping of the flexible tube and thus measure the critical velocity for a certain range of the mass-damping parameter. A new formulation of the added mass as a function of the void fraction was proposed. This formulation has a better agreement with the experimental results because it takes into account the reduction of the void fraction in the vicinity of the tubes in a rotated triangular tube array. The experimental data were used to validate the theoretical results of the quasi-steady model. The validity of the quasi-steady model for two-phase flow was confirmed by the good agreement between its results and the experimental data. The time delay parameter determined in the first stage of the project has improved significantly the theoretical results, especially for high void fractions (90%). However, the model could not be verified for void fractions lower or equal to 50% because of the limitation of the water pump capability. Further studies are consequently required to clarify this point. However, this model can be used to simulate the flow induced vibrations in steam generators' tube bundles as their most critical parts operate at high void fractions (≥ 60%). Having verified the quasi-steady model for high void fractions in two-phase flow, the third and final stage of the project was devoted to the development of a computer code for simulating flow induced vibrations of a steam generator tube subjected to fluidelastic and turbulence forces. This code was based on the ABAQUS finite elements code for solving the equation of motion of the fluid-structure system, and a development of a subroutine in which the fluid forces are calculated and applied to the tube. (Abstract shortened by UMI.)

Modelisation et simulation de pyrolyse de pneus usages dans des reacteurs de laboratoire et industriel

NASA Astrophysics Data System (ADS)

Lanteigne, Jean-Remi

The present thesis covers an applied study on tire pyrolysis. The main objective is to develop tools to allow predicting the production and the quality of oil from tire pyrolysis. The first research objective consisted in modelling the kinetics of tires pyrolysis in a reactor, namely an industrial rotary drum operating in batch mode. A literature review performed later demonstrated that almost all kinetics models developed to represent tire pyrolysis could not represent the actual industrial process with enough accuracy. Among the families of kinetics models for pyrolysis, three have been identified: models with one single global reaction, models with multiple combined parallel reactions, and models with multiple parallel and series reactions. It was observed that these models show limitations. In the models with one single global reaction and with multiple parallels reactions, the production of each individual pyrolytic product cannot be predicted, but only for combined volatiles. Morevoer, the mass term in the kinetics refers to the final char weight (Winfinity) that varies with pyrolysis conditions, which yields less robust models. Also, despite the fact that models with multiple parallels and series reactions can predict the rate of production for each pyrolysis product, the selectivities are determined for operating temperatures instead of real mass temperatures, giving models for which parameters tuning is not adequate when used at the industrial scale. A new kinetics model has been developed, allowing predicting the rate of production of noncondensable gas, oil, and char from tire pyrolysis. The novelty of this model is the consideration of intrinsic selectivities for each product as a function of temperature. This hypothesis has been assumed valid considering that in the industrial pyrolysis process, pyrolysis kinetics is limiting. The developed model considers individual kinetics for each of the three pyrolytic products proportional to the global decomposition kinetics of pyrolysables. The simulation with data obtained in industrial operation showed the robustness of the model to predict with accuracy in transient regime, tires pyrolysis, with the help of model parameters obtained at laboratory scale, namely in regards of the trigger of production, the residence time of tires (dynamic production) and the amount of oil produced (cumulative yield). It is a novel way to model pyrolysis that could be extrapolated to new waste materials. The second objective of this doctoral research was to determine the evolution of specific tires specific heat during pyrolysis and the enthalpy of pyrolysis. The origin of this objective comes from a primary contradiction. With few exceptions, it is acknowledged that organic materials pyrolysis is globally an endothermic phenomenon. At the opposite, all experiments led with laboratory apparatuses such as DSC (Differential Scanning Calorimetry) showed exothermic peaks during dynamic experiments (constant heating rate). It has been confirmed by results obtained at the industrial scale, where no sign of exothermicity has been observed. The Hess Law has also confirmed these results, that globally, pyrolysis is indeed a completely endothermic process. An accurate energy balance is required to predict mass temperature during pyrolysis, this parameter being unbindable from kinetics. An advanced investigation of char first allowed demonstrating that specific heat of solids during pyrolysis decreases with increasing temperature until the weight loss peak is reached, around 400°C, and then starts increasing again. This observation, combined with the fact that the sample loses weight during pyrolysis is considered as the major cause of the apparition of an exothermic peak in laboratory scale experiments. That is, the control system of these apparatuses generates a bias and an unavoidable overheat of the samples producing this exothermic behavior. It would thus be an artifact. On the base of new data on the evolution of global specific heat during pyrolysis, a model of the energy balance has been developed at the industrial scale to determine the enthalpy of pyrolysis. The simulation has shown that a major part of the heat transferred to the pyrolized mass would make its temperature increase. Next, an enthalpy of pyrolysis dependent of weight loss was obtained. Finally, two other terms of enthalpy have been found, namely an enthalpy for the breakage of sulfur bridges and an enthalpy for the stabilization of char when conversion approaches completion. This research will have allowed establishing a novel general methodology to determine the enthalpy of pyrolysis. More particularly, new clarifications hasve been obtained in regards to the evolution of specific heat of solids during pyrolysis and new enthalpies of pyrolysis, all endothermic, could be obtained, in agreement with the theoretical expectations. The third research objective concerned the behavior of sulfur during tires pyrolysis. With as a premise that sulfur is an intrinsic contaminant of many types of waste, it is critical to clarify its fate during pyrolysis, in the present case for waste tires. It has been observed in the literature that some quantitative analyses had been presented, but generally, the mechanisms for the distribution of sulfur within the pyrolytic products remain unclear. Thus, it was then not possible to predict the transfer of sulfur to each of the tire pyrolysis products. The results taken form literature have been complemented with a series of TGA experiments followed by complete elemental analyses of the residual solids. Mass balances have been performed in order to characterize the distribution of elements within the three products (noncondensable gas, oil, and char). A novel parameter has been created during this research: the sulfur loss selectivity. This intrinsic selectivity is a prediction of the distribution of sulfur within the pyrolysis products as a function of temperature. Three phenomena has been identified that could affect the sulfur loss selectivity. First, the natural devolatilization of sulfur due to pyrolysis. Next, the sulfur devolatilization due to the desulfurization of the solid matrix by hydrogen and finally, the clustering of sulfur in the solid state due to metal sulfidation (zinc and iron). The results have shown that this selectivity reach a limit value of 1 when pyrolysis is limited by the kinetics and in the absence of metal. When the mass transfer is limiting at low temperature (<500°C) the selectivity will be greater than 1. At a temperature over 350°C with the presence of metals, the selectivity will be lower than 1. It is a useful tool for industrial pyrolysis processes, being a novel indicator for the distribution of contaminants during the pyrolysis of waste. A better comprehension of these mechanisms allows elaborating a better strategy when designing these industrial processes. For example, in light of this research, it could be preferable to pre-treat the tires at lower temperature to eliminate a significant part of sulfur before pyrolyzing them at high temperature. The resulting pyrolytic products would then necessitate a lighter purification post-treatment, being more efficient and more economical.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times canmore » exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the contribution to the detector by a single neutron in each of the 47 energy groups used. Thus for the single existing symmetric pitch in the pools, where the vertical and horizontal separations are equal, only 6 sets of transfer functions are required. For the two asymmetrical pitches, nine sets of transfer functions are stored. In addition, source spectra at burnups ranging from 4000 to 20000 MWd/t and cooling times up to 40 years are stored. These source terms were established based on CANDU 37-rod fuel that is very similar to the Atucha fuel. Linear interpolation is used by the software for both burnup and cooling time to establish source terms at any intermediate condition. Using the burnup, cooling time and initial enrichment of the surrounding assemblies a set of source strengths in the 47-group structure for each of the 36 assemblies is established and multiplied group-wise with the appropriate transfer function set. The grand total over the 47 groups for all 36 assemblies is the predicted signal at the detector. The software was initially calibrated against a set of typically 5-6 measurements chosen from among the measured data at each level of the six pools and calibration factors were established. The set used for calibration is chosen such that it is fairly representative of the range of spent fuel assembly characteristics present in each level. Once established, these calibration factors can be repeatedly used for verification purposes. Recalibration will be required if the hardware or pool configurations has changed. It will also be required if a long enough time has elapsed since they were established thus making a cooling time correction necessary. The objective of the inspection is to detect missing fuel from one or more nearest neighbors of the detector. During the verification mode of the software, the predicted and measured signals are compared and the inspector is alerted if the difference between the two signals is beyond a set tolerance limit. Based on the uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.« less