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Technical Brief  
Jordan G. Gilbert, Scott B. Nokleby and Ed Waller
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036354
Inspections of pressure tubes in CANDU reactors are a key part of maintaining safe operating conditions. The current inspection system, the Channel Inspection and Gauging Apparatus for Reactors (CIGAR), performs the job well but is limited by the fact that it can only inspect one channel at a time. A multidisciplinary team is currently developing a novel robotic inspection system. As part of this work, a Monte Carlo N-Particle (MCNP) model has been developed in order to predict the dose rates that the improved inspection system will be exposed to and, from this, predict the component lifetime. This MCNP model will be capable of predicting in-core dose rates at any location within the reactor, and as such could be used for other situations where the in-core dose rate needs to be know.Based on estimates from this model, it is expected that at 7 days after shutdown the improved inspection system could survive in core for approximately 7 hours, providing it uses a tungsten shield 2:5 cm in thickness around the integrated circuit components. This is expected to be sufficient to perform a single inspection. 1 Introduction and Background
TOPICS: Pressure, Inspection, Particulate matter, Simulation, Robotics, Integrated circuits, Teams, Tungsten
research-article  
Govert de with
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036322
Fly ash is widely used as a supplementary cementitious material in the production of cement and concrete, and improves durability and strength of the concrete. However, as for all materials of mineral origin fly ash is a source for natural radioactivity; hence, its need for responsible use. The aim of this study is to investigate the radiation impact from fly ash as an additive to concrete compared against concrete without fly ash. For this purpose eight concrete mixtures are experimentally tested, followed by a computation of the radiation dose when used as bulk material in building constructions. The results demonstrate an increase in the total radiation dose from around 0.8 mSv with no fly ash up to 0.92 mSv when fly ash is used. The increase mostly comes from external radiation, while the radon exhalation factor reduces and sometimes even reduces the radon dose despite the higher radium content. The work has demonstrated that the impact from fly ash on the radiation exposure is limited when applied as a supplementary cementitious material. At the same time fly ash provides real benefits to the quality and durability of the concrete. For this reason exemption strategies for such applications should be developed.
TOPICS: Concretes, Radiation (Physics), Fly ash, Durability, Computation, Radioactivity, Radium, Cements (Adhesives), Bulk solids, Minerals
Technical Brief  
Eyal Peri, Adi Abraham, Tuvia Kravchik, Marcelo Weinstein, Dani Sattinger and Omer Pelled
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036137
The ICRP in its statement on tissue reaction from April 2011 (?1) recommended to reduce the annual dose limit to the lens of the eye from 150 mSv to 20 mSv (averaged over five years), with no single year dose exceeding 50 mSv. IAEA TECDOC 1731 (?2) and ISO 15382 (?3) were published as guidelines to the implementation of the new limits for occupational radiation. The most accurate way to determine the dose to the lens of the eye is to use a dosimeter that is designed and calibrated for measuring Hp(3), but it will take some time (probably a couple of years) until such dosimetric system could be validated and implemented in routine monitoring. Meanwhile, in some cases, there is a need to estimate Hp(3), especially for retrospective reconstruction of dose to the lens of the eye. Therefore, in the following years, the measured values of Hp(10) and or Hp(0.07) can be used to calculate a conservative value for Hp(3). The present paper discusses a new, more accurate and less conservative way to estimate Hp(3) using Hp(0.07) and Hp(10) quantities. This new method could be used also in reporting historical personal lens of the eye doses, when Hp(3) dosimeters were not used, and in some cases it could reduce the need to use special Hp(3) dosimeters in the future.
TOPICS: Lenses (Optics), Dosimeters, Radiation (Physics), Biological tissues
research-article  
Abhishek Kumar Srivastava, Rakesh Chouhan, Ananta Borgohain, S. S. Jana, N. K. Maheshwari, D.S. Pilkhwal, A. Rama Rao, K. N. Hareendhran, S. Chowdhury, K. B. Modi, S. K. Raut and S. C. Parida
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036027
Conceptual Molten Salt Breeder Reactor (MSBR) is under development in BARC with longterm objective of utilizing abundant thorium available in India. It is based on molten salts, which acts as fuel, blanket and coolant for the reactor. LiF-ThF4 (77.6-22.4 % mole) is proposed as a blanket salt for Indian MSBR. A laboratory scale molten salt natural circulation loop named, Molten Active Fluoride salt Loop (MAFL) has been setup for thermal-hydraulic, material compatibility and chemistry control studies. Various steady states and transient experiments have been performed in the operating temperature range of 600oC to 750oC. The loop operates in the power range of 250 W to 550 W. Steady state correlation given for natural circulation flow in a loop is compared with the steady state experimental data. The Reynolds number was found to in the range of 57 to 114. CFD simulation has also been performed for the same using OpenFOAM code and the results are compared with the experimental data generated in the loop. It has been found that the predictions of OpenFOAM are in good agreement with the experimental data. In this paper, features of the loop, its construction, the experimental and theoretical studies performed are discussed in detail.
TOPICS: Breeder reactors, Steady state, Operating temperature, Flow (Dynamics), Fuels, Reynolds number, Simulation, Construction, Coolants, Transients (Dynamics), Computational fluid dynamics, Chemistry
research-article  
Kurt E. Harris, Kevin J. Schillo, Yayu M. Hew, Akansha Kumar and Steven D. Howe
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035974
In NASA's Design Reference Architecture 5.0 (DRA 5.0), fission surface power systems (FSPS) are described as “enabling for the human exploration of Mars”. This study investigates the design of a power conversion system (PCS) based on supercritical CO2 (S-CO2) Brayton configurations for a growing Martian colony. Various configurations utilizing regeneration, intercooling, and reheating are analyzed. A model to estimate the mass of the PCS is developed and used to obtain a realistic mass-optimized configuration. This mass model is conservative, being based on simple concentric tube counterflow heat exchangers and published data regarding turbomachinery masses. For load following and redundancy purposes, the FSPS consists of three 333 kWe reactors and PCS to provide a total of 1MWe for 15 years. The optimal configuration is a S-CO2 Brayton cycle with 60% regeneration and two stages of intercooling. Analyses are mostly performed in MATLAB, with certain data provided by a COMSOL model of part of a low-enriched uranium (LEU) ceramic metallic (CERMET) reactor core.
TOPICS: Nuclear fission, Optimization, Power conversion systems, Brayton cycle, Supercritical carbon dioxide, Design, Heat exchangers, Matlab, Turbomachinery, Uranium, Power systems (Machinery), Ceramics, Cermets, Stress, Redundancy (Engineering)
research-article  
Xiaoming Chai, Xiaolan Tu, Wei Lu, Zongjian Lu, Dong Yao, Qing Li and Wenbin Wu
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035934
Due to powerful geometry treatment capability, Method Of Characteristics (MOC) currently becomes one of best method to solve neutron transport equation. In MOC method, boundary condition treatment, complex geometry representation, and advanced acceleration method are the key techniques to develop a powerful MOC code to solve complex problem. In this paper, we developed a powerful MOC module, which can treat different boundary conditions with two methods. For problems with special border shapes and boundary condition, such as rectangle, 1/8 of square, hexagon, 1/6 of hexagon problems with reflection, rotation, and translation boundary condition, the MOC module adopts periodic tracking method. For problems with general border shapes, the MOC module use ray prolongation method. Meanwhile, graphic user interface based on CAD software is developed to generate the geometry description file. The geometry and mesh can be described and modified correctly and fast. In order to accelerate the MOC transport calculation, the Generalized Coarse Mesh Finite Differential (GCMFD) is used, which can use irregular coarse mesh diffusion method to accelerate the transport equation. The MOC module was incorporated into advanced neutronics lattice code KYLIN-2, which developed by Nuclear Power Institute of China (NPIC) and used to simulate the assembly of current PWR reactor and advanced reactors. The numerical results show that the KYLIN-2 code can be used to calculate 2D neutron transport problems accurately and fast. In future, the KYLIN-2 code will be released and gradually become the main neutron transport lattice code in NPIC.
TOPICS: Rotation, Diffusion (Physics), Neutrons, Manufacturing, Reflection, Computer-aided design, Boundary-value problems, China, Computer software, Geometry, Nuclear power, Shapes, User interfaces, Pressurized water reactors
research-article  
Ella Israeli and Erez Gilad
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035883
Novel genetic algorithms are developed by using state of-the-art selection and crossover operators, e.g., rank selection or tournament selection instead of the traditional roulette (fitness proportionate) selection operator and novel crossover and mutation operators by considering the chromosomes as permutations (which is a specific feature of the loading pattern problem). The algorithm is implemented and applied to a representative model of a modern PWR core and for a single objective fitness function, i.e., k_eff. The results obtained for some reference cases using this setup are excellent and are obtained by utilizing a tournament selection operator with a linear ranking selection probability method, and a new geometric crossover operator that allows for geometrical swaps, rather than random, of genes segments between the chromosomes and control the sizes of the swapped segments. Finally, the effect of boundary conditions on the symmetry of the obtained best solutions is studied and the validity of the "symmetric loading patterns" assumption is tested.
TOPICS: Genetic algorithms, Fuel management, Probability, Pressurized water reactors, Algorithms, Boundary-value problems
research-article  
X. Gaus-Liu and A. Miassoedov
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035853
This study investigates experimentally the heat transfer characters of a volumetrically-heated melt pool in LWR lower plenum under different cooling boundary conditions both in 2D and 3D vessel geometries and with simulant melts with and without potential of crust formation. A survey of existing heat transfer correlations based on individual experimental definitions is firstly given. The inconsistency in parameter definitions in Nu-Ra correlations is addressed. The general difference of upward heat transfer behavior in term of Nu depending on the existence of crust is discussed and explained. In several serials of LIVE3D and LIVE2D experiments, in which different combination of external cooling and top cooling were performed, and both simulants with crust formation and without crust formation were used, the influence of cooling boundary conditions, the vessel geometry and the simulants on the overall upwards and downwards heat transfer as well as on the melt temperature and heat flux distribution have been analyzed. This paper provides some explanations about the discrepant among the exiting heat transfer correlations and recommends most suitable descriptions of melt pool heat transfer under different accident management scenarios.
TOPICS: Boundary-value problems, Geometry, Vessels, Heat transfer, Light water reactors, Cooling, Accident management, Heat flux, Temperature
Technical Brief  
R. Reuven, A. M. Bolind, N. Haneklaus, C. Cionea, C. Andreades, G. Buster, P. Hosemann and P. F. Peterson
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035725
This study suggests a new approach to diffusion bonding (DB) 316L stainless steel: a low-pressure procedure that includes a nickel interlayer. In this approach, relatively lower pressure is applied to the sample before the DB process, in contrast to the usual approach in which higher pressure is applied during the DB process. This new procedure was tested on mock-up 316L stainless steel tube-to-tubesheet joints, which simulated similar joints in coiled-tube heat-exchanger applications. This study confirms that the new procedure meets the overall success criteria, namely, a pull-out force exceeding the force required for tube rupture. It also shows that the DB joint is improved by the use of a Ni interlayer; the joint strength increased by approximately 33% for a 0.25 µm Ni interlayer and by approximately 18% for a 5 µm Ni interlayer. The joint cross-sections were qualitatively examined using optical microscopy and scanning electron microscopy; the observations suggest that only portions of the interface were diffusion bonded, as a result of the low-pressure procedure and the surface roughness (due to the sample fabrication). The portions that were diffusion bonded, though, were sound, as characterized by the fact that the steel grains grew through the interface line to create a continuous metallographic structure.
TOPICS: Pressure, Diffusion bonding (Metals), Press fits, Stainless steel, Diffusion (Physics), Nickel, Steel, Manufacturing, Surface roughness, Cross section (Physics), Heat exchangers, Scanning electron microscopy, Optical microscopy, Rupture
research-article  
Brandys Irad, Ornai David and Ronen Yigal
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035692
Standards, guidelines, manuals and researches, refer mainly to the required protection of a nuclear power plant (NPP) containment structure (where the reactor's vessel is located) against different internal and external extreme events. However, there is no consideration regarding the man-made extreme event of external explosion resulting from air bomb or cruise missile. A novel Integrated Blast Resistance Model (IBRM) of NPP's reinforced concrete (RC) auxiliary facilities due to an external above ground explosion based on two components is suggested. The first is structural dynamic response analysis to the positive phase of an external explosion based on the single degree of freedom (SDOF) method combined with spalling and breaching empirical correlations. The second is in-structure shock analysis, resulted from direct-induced ground shock and air-induced ground shock. As a case study, the resistance of Westinghouse commercial NPP AP1000 control room, including a representative equipment, to an external explosion of Scud B-100 at various stand-off distances ranging from 250m (far range) till contact, was analyzed. The structure's damage level is based on its front wall supports' angle of rotation and the ductility ratio. Due to the lack of specific structural damage demands and equipment's dynamic capacities, common protective structures standards and manuals are used while requiring that no spalling nor breaching shall occur as well as remaining in the elastic regime. The IBRM may be used in wider researches concerning other NPP's auxiliary facilities and systems based upon their specifications.
TOPICS: Nuclear power stations, Explosions, Shock (Mechanics), Damage, Rotation, Vessels, Containment, Bombs, Structural dynamics, Degrees of freedom, Missiles, Reinforced concrete, Control rooms, Ductility
research-article  
Liel Ishay, Ulrich Bieder, Gennady Ziskind and Alexander Rashkovan
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035693
Knowledge of the Nuclear Power Plants (NPPs) containment atmosphere composition in the course of a severe accident is crucial for the effective design and positioning of the hydrogen explosion countermeasures. This composition strongly depends on containment flows which may include turbulent jet mixing in the presence of buoyancy, jet impingement onto the stratified layer, stable stratification layer erosion, steam condensation on the walls of the containment, condensation by emergency spray systems and other processes. Thus, in modelling of containment flows, it is essential to predict correctly these effects. In particular, a proper prediction of the turbulent jet behaviour before it reaches the stably stratified layer is critical for the correct prediction of its mixing and impingement. Accordingly, validation study is presented for free neutral and buoyancy-affected turbulent jets, based on well-known experimental results from the literature. This study allows for the choice of a proper turbulence model to be applied for containment flow simulations. Furthermore, the jet behaviour strongly depends on the issuing geometry. A comparative study of erosion process for the conditions similar to the ones of International Benchmark Exercise (IBE-3) is presented for different jet nozzle shapes.
TOPICS: Flow (Dynamics), Modeling, Nuclear power stations, Containment, Turbulence, Buoyancy, Condensation, Erosion, Flow simulation, Shapes, Steam, Nozzles, Sprays, Geometry, Hydrogen, Explosions, Jets, Accidents, Design, Emergencies, Countermeasures
Technical Brief  
A. S. Schneider and N. Yair
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035565
Various questions can be examined when discussing safety in general. Among these some key issues are the attitude towards risk and its acceptance, the ways of identifying, analyzing and quantifying risks, and societal factors and public opinion towards risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these we mention the defense-in-depth approach, the design-basis-accident (DBA) and beyond-design-basis-accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making.
TOPICS: Safety, Nuclear reactors, Accidents, Design, Decision making, Defense industry, Risk
Technical Brief  
A. Biton, Y. Soffer and R. Freud
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035568
Nuclear facilities and in particular nuclear reactors are designed to withstand ground acceleration due to earthquake and to maintain the structures, systems and components (SSCs) safety function. The three main safety functions are shutdown, removal of residual heat and indications of the first two. Achieving these three functions will be called a "Safe path". The IRR2 has numerous diverse "safe paths" to ensure the availability of these three safety function. Each "safe path" has an associated level of resistance to ground acceleration. A Project for increasing the robustness of equipment to ground acceleration was initiated in order to improve the safety of the reactor and to comply with regulatory guidelines.
TOPICS: Safety, Robustness, Heat, Earthquakes, Nuclear reactors, Nuclear power stations
research-article  
Graham Steeves and William Cook
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035549
Corrosion behaviour of Inconel 625 and Incoloy 800H, two of the candidate fuel cladding materials for Canadian supercritical water (SCW) reactor designs, were evaluated by exposing the metals to SCW in UNB's SCW flow loop. A series of experiments were conducted over a range of temperatures between 400oC and 600oC and the corrosion rates were evaluated as the weight change of the materials over the exposure time (typical experiments measured the weight change at intervals of 100, 250, and 500 hours, with some longer term exposures included). SEM, EDX, and TEM techniques were used to examine and quantify the oxide films formed during exposure and the corrosion mechanisms occurring on the candidate metals. Data from in-house experiments were used to create an empirical kinetic equation for each material that was then compared to literature values of weight change. Dissolved oxygen concentrations varied between experimental sets, but for simplicity were ignored since the effect of dissolved oxygen has been demonstrated to be a minor secondary effect. Activation energies for the alloys were determined with Inconel 625 and Incoloy 800H showing a distinct difference between the low-temperature electrochemical corrosion mechanism and direct high-temperature chemical oxidation. The results were modelled using these separate effects showing dependence on the bulk density and dielectric constant of the SCW through the hydrogen ion concentration.
TOPICS: Alloys, Corrosion, Supercritical water reactors, Weight (Mass), Metals, Oxygen, Water, High temperature, Low temperature, Hydrogen, oxidation, Density, Flow (Dynamics), Temperature, Fuels, Electrolytic corrosion, Cladding systems (Building)
research-article  
Emiliya L. Georgieva, Yavor D. Dinkov and Kostadin Ivanov
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035550
The aim of this paper is to summarize authors' experience in adaptation of an existing plant-specific VVER-1000/V320 model for simulation of a rare example of a Kalinin 3 NPP transient of 'Switching-off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power' with an asymmetric core configuration. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. Simulation results concerning fuel assembly power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system. Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superficial description of the reference unit. In such a case, an approach based on a 'generic' V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion. 1) individual characteristics of all the main circulation pumps and the reactor coolant loops; 2) variations in fuel assembly characteristics ; 3) dynamic response of instrumentation and control systems; 4) balance-of-plant equipment, instrumentation and control. Above requirements impose a difficult task to comply with. Nevertheless, any individual nuclear power unit is supposed to maintain a detailed design data base and data requirements for plant specific model development should be considered.
TOPICS: Control systems, Fuels, Manufacturing, Simulation, Transients (Dynamics), Design, Instrumentation, Pumps, Databases, Dynamic response, Nuclear reactor coolants, Model development, Monitoring systems, Nuclear power, Nuclear power stations, Simulation results, Steam
research-article  
Gabriela Bar-Nes, Yael Peled, Zorik Shamish and Amnon Katz
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035415
The effect of incorporation of pozzolanic additives on the immobilization of Cesium and Strontium ions in cementitious pastes was investigated. Pastes containing Portland cement together with ground granulated blast furnace slag (50%, 75%), metakaolin (10%, 20%) or silica fume (20%), either in its densified or raw form, were prepared. The transport properties of the immobilized ions through the paste were evaluated using leaching tests. Single differential thermal analysis was used to estimate the extent of the pozzolanic reaction and the pozzolanic reactivity of the different formulations. For the Strontium ion, the best immobilization system was the 20% Raw Silica Fume (RSF) paste, characterized by the highest relative pozzolanity. For Cesium ions however, the most effective additive was the Densified Silica Fume (DSF), reducing the apparent diffusion coefficient by two orders of magnitude compared to the unblended paste.
TOPICS: Cements (Adhesives), Strontium, Ions, Blast furnaces, Slags, Thermal analysis, Diffusion (Physics)
research-article  
Avraham Dody, Ravid Rosenzweig, Ran Calvo and Eyal Shalev
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035405
Two main natural processes were studied in the Yamin Plain in order to evaluate the thickness of a cover layer of a near-surface waste disposal facility, needed to prevent potential migration of radioactive contaminants to the biosphere. The first is the natural erosion rate of the cover layer, and the second is the infiltration depth during rain and runoff events. The erosion rate was studied by optical stimulation luminescence technique. It was found that during the last 14,000 years, the erosion rate was 0.3 mm/y. The infiltration depth assessment was based on water content measurements and numerical modeling. It shows that in the most extreme rain even having an equivalent rain of 84 mm, infiltration depth was limited to 4.5 m. Therefore, effective cover layer over 10,000 years, should be at least 7.5 m thick.
TOPICS: Waste disposal, Erosion, Water, Computer simulation, Luminescence
research-article  
Gaoming Ge, Tomohiko Ikegawa, Koji Nishida and Carey J. Simonson
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035335
Hitachi-GE developed a 300MWe class DMS (Double MS: Modular simplified & medium small reactor) concept and the DMS was originally designed for generating electricity only. In this study, the feasibility of a cogeneration DMS plant which supplies both electricity and heat is under investigation. The thermal performance of the DMS plant without or with low, medium or high temperature thermal utilization (TU) applications is evaluated by numerical simulations. The results show the electricity generated reduces as the heating requirement of TU application becomes higher. Furthermore, the economic performance of the cogeneration DMS plant is compared with another two integrated systems: (i) DMS plus electric boilers and (ii) DMS plus natural gas boilers, for those three TU applications in Canada. Results illustrate the DMS plus natural gas boilers system is most economic if there is no carbon tax, but with high CO2 emissions (up to 180 Ktons per year). The cogeneration plant performs best as the carbon tax increases up to 40$/ton. The cogeneration DMS plant is a promising scheme to supply both electricity and heat simultaneously in the economic-environmental point of view.
TOPICS: Combined heat and power, Nuclear reactors, Thermoeconomics, Boilers, Carbon, Natural gas, Heat, Computer simulation, Electric power generation, Carbon dioxide, Heating, Emissions, Integrated systems, High temperature, Cogeneration plants
research-article  
Yan Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4032312
In modeling and simulation, model-form uncertainty arises from the lack of knowledge and simplification during modeling process and numerical treatment for ease of computation. Traditional uncertainty quantification approaches are based on assumptions of stochasticity in real, reciprocal, or functional spaces to make them computationally tractable. This makes the prediction of important quantities of interest such as rare events difficult. In this paper, a new approach to capture model-form uncertainty is proposed. It is based on fractional calculus, and its flexibility allows us to model a family of non-Gaussian processes, which provides a more generic description of the physical world. A generalized fractional Fokker-Planck equation (fFPE) is used to describe the drift-diffusion processes under long-range correlations and memory effects. A new model calibration approach based on the maximum mutual information is proposed to reduce model-form uncertainty, where an optimization procedure is taken.

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