bluuebunny/crossref_metadata_2025_split huggingface dataset:
CANDU reactors with reactor grade plutonium/thorium carbide fuel
Authors: Şahin S. , Khan M. J. , Ahmed R. 
Published: December 2013

Abstract
Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k∞,0 = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are ∼2.7, 8.4, and 15 years and with burn ups of ∼72000, 222000 and 366000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced.
Dosimetric characteristics of three new design 125I brachytherapy sources
Authors: Khanmohammadi Z. , Sadeghi M. 
Published: December 2013

Abstract
For roughly 25 years, 125I sources have been used in the treatment of various malignant diseases such as prostate cancer. Three new brachytherapy sources, IR01-125I, IR02-125I and IR03-125I, have been developed and are designed for permanent implant application. The Monte Carlo radiation transport code version MCNP 5 was used to calculate the dosimetry parameters around the sources in accordance with the updated report of the American Association of Physicists in Medicine (AAPM), Task Group No. 43. For each source, the dose rate constant Λ, the radial dose function gL(r), and the anisotropy function F(r, θ), were obtained. The results indicated a dose rate constant of 0.932 ± 0.01, 0.934 ± 0.01 and 0.939 ± 0.01 Gy h−1 U−1 for the IR01-125I, IR02-125I and IR03-125I sources respectively. With the goal of determining an optimal design for a 125I source, each seed's parameters were compared with other seeds. In this study, the optimal source IR03-125I provides the most isotropic dose distribution in water. Finally, the results for optimal source were compared with published results for those of other commercial sources.
Novel dose calculation and characterization of 32P intravascular brachytherapy stent source
Authors: Sadeghi M. , Kiavar O. , Saidi P. , Fatehi R. 
Published: December 2013

Abstract
Derived from AAPM task group No. 60/149 protocol, applicable in treatment planning In this study, the two-dimensional dose distributions in water for a 32P intravascular brachytherapy stent have been calculated. The pure beta emitter source 32P which has been coated on Palmaz-Schatz stent is discussed. The dosimetric parameters required by the AAPM TG-60/149 formalism are discussed and calculated. Version 5 of the (MCNP) Monte Carlo radiation transport code was used to calculate the dosimetry parameters around the source. The Monte Carlo calculated dose rate at the reference point is found to be 2.8 Gy/μCi. Also in this study, the geometry function, G(r,θ), radial dose function, g(r), and the anisotropy function, F(r,θ), have been calculated at distances from 1.8 to 9 mm. The results of these calculations have been compared with other published calculated and measured values for an actual same source. High dose variants were visible near the 32P stent surface, but these values decreased with depth in water rapidly. There is an acceptable agreement between the calculated data in this study and other published data for the same source, which validate our simulations method.
Burn up extension in a PBMR-400 full core using weapon grade plutonium fuel mixed with thorium
Authors: Acir A. , Coşkun H. 
Published: December 2013

Abstract
In this study, the time dependent criticality and burn up analyses for a full core model of Pebble Bed Modular Reactors, such as PBMR-400, are performed using weapon grade plutonium mixed with thorium as an alternative to enriched UO2 fuel. Neutronic calculations are performed with the MCNP and MONTEBURNS 2.0 using ENDF/B-V nuclear data library. 9.6% enriched UO2 gives an effective multiplication factor of keff ≈ 1.2395, which can be achieved by 18% weapon grade plutonium +82% ThO2 fuel mixture. Thorium mixed fuel with 18% weapon grade plutonium fuel would allow reactor operation periods ∼7 years without fuel change with burn up levels of 205000 MW.D/t, for end of life keff set equal to 1.02.
Stress analyses for reactor pressure vessels by the example of a product line '69 Boiling Water Reactor
Authors: Mkrtchyan L. , Schau H. , Wolf W. , Holzer W. , Wernicke R. , Trieglaff R. 
Published: December 2013

Abstract
The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions.
Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part I: key areas
Authors: Sinha N. K. , Raj B. 
Published: December 2013

Abstract
Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s–early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton® GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution.
Efficiency of emergency exercises
Authors: Zander N. , Sogalla M. 
Published: December 2013

Abstract
In order to cope with accidents beyond the design basis within German nuclear power plants which possibly lead to relevant radiological consequences, the utilities as well as the competent authorities exist emergency organisations. The efficiency, capacity for teamwork and preparedness of such organisations should be tested by regular, efficient exercise activities. Such activities can suitably be based on scenarios which provide challenging tasks for all units of the respective emergency organisation. Thus, the demonstration and further development of the efficiency of the respective organisational structures, including their ability to collaborate, is promoted.
Electronic situation report in a nuclear power plant
Authors: Wilbois T. , Haas H. , Poppe W. , Chaves F. , Ren Y. 
Published: December 2013

Abstract
During the last years the introduction of an electronic situation report system in nuclear power plants has been intensively discussed in the working group emergency staff coordinators of the VGB (‘Verband der Großkraftwerksbetreiber’). In parallel, many federal state authorities as well as the German Ministry of the Environment already successfully started to operate such a system. Based on a solution developed for the ministry for environment Baden-Württemberg, the specific requirements are adapted and extended for the operation in a nuclear power plant. The most important features are easy handling, open interfaces to the systems of the authorities, and flexibility for functional extensions. The advantages recognized after the first test stages led the nuclear power plants Biblis and Neckarwestheim to the decision to introduce such a system at the plant site. First progress and experience gained are reported from the perspective of the crisis management group of the power plant.
Nuclear emergency Exercise “BIBLIS 2008” – Joint radiological assessment for Hessen, Rheinland-Pfalz and Baden-Württemberg
Authors: Kraus G. 
Published: December 2013

Abstract
Off-site nuclear emergency preparedness has to deal with extensive scenarios. In consequence, radiologial assessment in such cases has to follow a transboundary approach. The joint evaluation and assessment of the radiological situation and the decision on emergency countermeasures after a nuclear accident in the NPP Biblis as required in the Basic Recommendations for Civil Protection was trained in a concerted excercise by Hessen, Rheinland-Pfalz und Baden-Württemberg in autumn 2008.
Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor
Authors: Horhoianu G. , Serbanel M. , Diaconu C. 
Published: December 2013

Abstract
Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavodâ-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described.
On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures
Authors: Jäger W. , Espinoza V. H. Sánchez 
Published: December 2013

Abstract
This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment.
PCA-based ANN approach to leak classification in the main pipes of VVER-1000
Authors: Hadad K. , Jabbari M. , Tabadar Z. , Hashemi-Tilehnoee M. 
Published: December 2013

Abstract
This paper presents a neural network based fault diagnosing approach which allows dynamic crack and leaks fault identification. The method utilizes the Principal Component Analysis (PCA) technique to reduce the problem dimension. Such a dimension reduction approach leads to faster diagnosing and allows a better graphic presentation of the results. To show the effectiveness of the proposed approach, two methodologies are used to train the neural network (NN). At first, a training matrix composed of 14 variables is used to train a Multilayer Perceptron neural network (MLP) with Resilient Backpropagation (RBP) algorithm. Employing the proposed method, a more accurate and simpler network is designed where the input size is reduced from 14 to 6 variables for training the NN. In short, the application of PCA highly reduces the network topology and allows employing more efficient training algorithms. The accuracy, generalization ability, and reliability of the designed networks are verified using 10 simulated events data from a VVER-1000 simulation using DINAMIKA-97 code. Noise is added to the data to evaluate the robustness of the method and the method again shows to be effective and powerful.
Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes
Authors: Kozmenkov Y. , Rohde U. , Baranaev Y. , Glebov A. 
Published: December 2013

Abstract
RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached.
The application of equilibrium models to incidence situations using the example of the exposure pathway human milk
Authors: Steiner M. , Karcher K. , Noßke D. 
Published: December 2013

Abstract
The radiation exposure after a short-term release of radioactive substances is often calculated assuming equilibrium conditions. An example is that of the German Incident Calculation Bases for nuclear power plants with pressurized water reactors. Here, the contamination of human milk is calculated using transfer factors. Applying this equilibrium model to incident situations raises the question whether baby's radiation exposure is adequately assessed. This contribution shows that compliance with the relevant dose limits of paragraph 49 of the German Radiation Protection Ordinance is ensured for design basis accidents on the assumption that the hypothetical breastfeeding period starts at the beginning of the activity release. Comparative analyses were performed against the biokinetic models applied by ICRP for radiation protection purposes, taking the reference nuclides 137Cs, 90Sr, 131I, 241Am and long-lived plutonium isotopes as examples.
The criticality calculations for one-speed neutrons in a reflected slab with anisotropic scattering using the modified UN method
Authors: Öztürk H. 
Published: December 2013

Abstract
The reflected critical slab problem is studied using Chebyshev polynomials of second kind in one-speed neutron transport theory. The variation of the critical thickness of the medium is investigated both in the case of linearly anisotropic scattering and reflecting boundary conditions. The critical half-thicknesses of the slab are calculated for selected values of the reflection coefficient and the anisotropy parameter. Numerical results obtained for the critical half-thickness are identical with the ones obtained by the PN method and the ones available in literature.
Remarks on boiling water reactor stability analysis – part 1: theory and application of bifurcation analysis
Authors: Lange C. , Hennig D. , Hurtado A. , Schuster R. , Lukas B. , Aguirre C. 
Published: December 2013

Abstract
Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants.
Development of a 1 D hybrid HTC model using CFD simulations for the analysis of direct contact condensation as the driving force for water hammers
Authors: Ceuca C. S. , Macián-Juan R. 
Published: December 2013

Abstract
A Hybrid Heat Transfer Coefficient module has been developed based on two Surface Renewal Theory models using CFD simulations. The validation of the model has been done on a meso-scale computational grid for CFD simulations and on a macro-scale computational grid for System Code analysis. The CFD simulation was performed for a stratified co-current two phase flow between saturated steam and sub-cooled water while the System Code analysis was performed for a Condensation Induced Water Hammer experiment.
Effect of gamma-radiation on sorption and precipitation of radionuclides
Authors: Kepák F. 
Published: December 2013

Abstract
The effect of gamma radiation on sorption of 85Sr2+ on trace manganese dioxide and on precipitation of 131I– with Ag+ ions has been studied. The transition of both radionuclides into the solid phase by sorption and precipitation was found by measurement of the self-diffusion coefficients of 85Sr2+ and of 131I–. The gamma radiation had negative effect on sorption and precipitation and also affected the ionic state of 85Sr2+ and of 131I–.
Automatic loading pattern optimization tool for Loviisa VVER-440 reactors
Authors: Kuopanportti J. 
Published: December 2013

Abstract
An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies’ operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution.
Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code
Authors: György H. , Trosztel I. 
Published: December 2013

Abstract
Severe accident – if no mitigation action is taken – leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor.
Contribution of the number of measured data to calculation uncertainty in the worth of VVER control rods
Authors: Sumarokova A. , Shishkov L. 
Published: December 2013

Abstract
During reactor start-up and operation, one of the most important parameters, which determine reactor safety, are the effectiveness of reactor shutdown system and the worth of working group control rods. The assessment of the calculation uncertainty of these functional is also important for the design of reactor loadings. The paper considers methods for estimating uncertainties by comparing calculated and measured data. The paper shows that the distribution of the deviations of calculated (BIPR-7A) and measured data is close to normal, and the characteristics of the distribution (average and standard deviations), which are determined for a final sample, are close to the parameters. It was shown, that the worth and effectiveness of control rod groups can be assessed by using the characteristics of finite samples.