bluuebunny/crossref_metadata_2025 huggingface dataset:
Some affine solutions for CANDLE burn-up waves in 1D-geometry
Authors: Seifritz W.
Published: December 2013

Using the method of “affine solutions” three interesting phenomena in a CANDLE Burn-Up strategy were studied and were cast analytically into mathematical formlulae: a) the problem of two burn-up waves one behind the other (“enslavement principle”) b) the problem of the frontal collision of two burn-up waves (“Scattering”), and c) the problem of two identical burn-wave twins moving apart from each other after central ignition (“Breathing Out Phenomenon”).

To the knowledge of the author this is the first time where such affine problems of the CANDLE Burn-Up wave physics were solved.
Simulation of H2 distribution and combustion in LWR containments using Lumped Parameter Codes
Authors: Burkhardt J. , Klöcker M. , Brähler T. , Koch M. K.
Published: December 2013

This article focuses on the validation of the Lumped Parameter (LP) computer codes COCOSYS and ASTEC in the field of containment phenomena during severe accidents. The release of hydrogen (H2) as an open jet from the primary circuit into the containment might cause atmospheric H2-stratifications which in turn may result in an ignitable atmospheric mixture. If this mixture is ignited, the integrity of the containment can be endangered due to the pressure and temperature rise, resulting from the combustion process. The emphasis of the paper lies on these three phenomena referring to possible successive phases of a loss of coolant accident.
CFD modelling of insulation debris transport phenomena in water flow
Authors: Krepper E. , Cartland-Glover G. , Grahn A.
Published: December 2013

The investigation of insulation debris generation, transport and sedimentation becomes important with regard to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during all types of loss of coolant accidents. A joint research project on such questions is being performed in cooperation between the University of Applied Sciences Zittau/Görlitz and the Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description. While the experiments are performed at the University at Zittau/Görlitz, the theoretical modelling efforts are concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modelling are described and feasibility studies are presented.
Closed loop auto control system software for Miniature Neutron Source Reactors (MNSRs)
Authors: Iqbal M. , Qadir J. , Bhatti T. K. , Abbas Q. , Mirza S. M.
Published: December 2013

A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response.
Chebyshev polynomials expansion method for solving the one-dimensional transport equation in spherical geometry
Authors: Yaşa F.
Published: December 2013

Using certain well-known properties of Chebyshev polynomials, an efficient analytical approach to evaluate the Boltzmann equation is presented in one-dimensional spherical geometry. The procedure is based on the expansion of the angular flux in terms of a series of Chebyshev polynomials.
Development of a decommissioning strategy for the MR research reactor
Authors: Bylkin B. , Craig D. , Fecitt L. , Gorlinsky Y. , Harman N. , Jackson R. , Kolyadin V. , Lobach Y. , Pavlenko V.
Published: December 2013

A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and overall decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown.
Methods for thin foil thickness determination by using alpha spectroscopy
Authors: Hao L. C. , Nhon M. V. , Tao C. V.
Published: December 2013

In both, industry and scientific research, there is a demand for the determination of material thicknesses. In this work, two methods are described which are very simple, fast, effective and low-cost to determine the exactly thickness of thin foils (μm). Both methods are based on the energy loss of alpha particles when passing through the foil and the detection by high energy resolution alpha spectroscopy. The strengths and weakness of the two methods are discussed.
The time-dependent P1 model for the neutronics of multiplying systems: a review
Authors: Coppa G. G. M. , Dulla S. , Ravetto P.
Published: December 2013

The derivation of the time-dependent P1 model by the spherical harmonics expansion approach to solve the neutron transport equation is reviewed and its main features and physical characteristics are illustrated. The telegrapher's equation is presented and its formulation for multiplying systems is discussed. In particular, it is evidenced that the presence of a time derivative of the source terms in the telegrapher's equation implies the appearance of time derivatives of the flux and of the precursor concentrations. The role of these terms is evaluated and discussed.
The albedo problem for pure-quadratic anisotropic scattering with İnönü scattering
Authors: Türeci R. G.
Published: December 2013

In this study the half-space albedo problem has been investigated for highly anisotropic scattering. For this highly anisotropic scattering the İnönü scattering has been added to the pure-quadratic anisotropic scattering. The general solution of the one-speed neutron transport equation can be written as linear combination of the solutions which involve these scatterings. Hence, Case's eigenfunctions and the related semi-analytical methods can be used to get the albedo calculations. Some selected numerical data have been tabulated. To compare these results with the İnönü scattering values the linear interpolation technique has been used. Successful results have been achieved.
Pertinence of tests used for conformity assessments of fire resistance of nuclear power plant components – opinion of several institutions in Europe
Authors: Bruynooghe C. , Röwekamp M. , Hostikka S. , Richard J. , Maillet E. , Neugebauer W. , Zdeněk T.
Published: December 2013

Fire resistance of structures, systems and components (SSC) is widely identified as a fundamental issue for fire risk assessment at nuclear power plants (NPP). It is currently supported by several national or international ambitious experimental programs. Fire risk assessment at NPPs is performed under a series of hypothesis, e.g. the fire load. A problem will arise if tests used for conformity assessment do not appropriately represent the hypothesis selected during the fire risk analysis. In this case, SSCs may possibly be not meeting the expectations related to their fire resistance capacity supposed by the fire risk analysis. Ways of tackling this issue involve a regulatory approach and adapted requirements for components. While the widespread regulatory approach consists in seeking enveloping scenarios and conservative assumptions, a better knowledge of SSCs behaviour under for NPPs typical situations may be also desirable. The Institute for Energy of the European Commission's Joint Research Centre (JRC) at Petten (NL) launched a questionnaire with the aim of gathering opinions of regulatory bodies and their technical support organisations and manufacturers in Europe on this topic. Major goal was to estimate the relevance of the issue according to the perception of the specialists. The present paper collects the answers given by seven entities (AREVA, GRS, iBMB, IRSN, SJUB, SUEZ-TRACTEBEL, and VTT) and makes suggestions for addressing the issues identified by the participants.
Analysis of the processes in the target cooling system of the W7-X fusion experiment
Authors: Kaliatka T. , Kaliatka A. , Kačegavičius T. , Naujoks D.
Published: December 2013

Nuclear fusion has the potential to cover a major part of mankind's energy needs under the conditions of ever-increasing population and energy consumption. Wendelstein 7-X (W7-X) is a fusion experiment currently being built in Greifswald, which shall demonstrate the capability of the stellarator concept on the way to a fusion power plant. This paper presents analyses of processes in the target cooling system with the codes RELAP5 and ASTEC. The loss-of-coolant accident (LOCA) during the “baking” operation mode of W7-X (i.e. under the most critical conditions because of the high temperature and pressure applied) is analysed. The response of the W7-X “baking” and main cooling circuits and the amount of coolant discharged into the plasma vessel are investigated.
Performance analysis of 233U for fixed bed nuclear reactors
Authors: Şahin S. , Acır A. , Şahin H. M.
Published: December 2013

Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.
Application of the Chebyshev polynomial (T

N
and U

N
) approximation to reflected slab geometry in the neutron transport equation and computation of critical half thicknesses
Authors: Bülbül A. , Anlı F.
Published: December 2013

The transport of one-speed neutrons has been studied in isotropically scattering slab using ultraspherical polynomial approximation whose sub-cases are spherical harmonics approximation and Chebyshev polynomial approximation of first and second kind with reflective boundaries. An integral equation for the total flux has been derived and solved numerically for different values of c and reflection coefficients. In the solution, Marshak type vacuum boundary conditions were used. The critical thicknesses obtained by the Chebyshev polynomial method are in good agreement with literature values. Computations were made by using the TN, UN and PN approximations for comparison.
Analysis of fuel rod behaviour during limiting RIA in RBMK plants
Authors: Pabarcius R. , Kaliatka A. , Marao A.
Published: December 2013

Typically, reactivity initiated accidents for nuclear power plants with RBMK type reactors are classified as design basis accidents. However, for assessment of the real probability of these events, such kind of incidents should be considered more seriously and the same acceptance criteria as for operational occurrences should be applied to them. The spurious withdrawal of control rods, as limiting reactivity initiated accident (RIA), was analysed and results of the study are presented in this paper. The power distribution in fuel channels located around the affected control rods was calculated using the QUABOX/CUBBOX-HYCA code. Evaluating possible uncertainties it was found that linear power, i. e. heat generated per unit length of coolant channel, exceeds the acceptance criterion by a factor of 1.21. This paper presents the discussion how such a peak of power influences the behaviour of RBMK-1500 reactor fuel rods. For this purpose, the FEMAXI-6 code was applied. Results of the analysis show, that after increasing and subsequently decreasing of the linear power during an accident most parameters (pressure inside fuel rods, radial displacements, fuel clad gaps, etc.) return to their initial values and no phenomena occurred that would preclude the resumption of normal operation after termination of such an event.
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
Authors: Horhoianu G. , Ionescu D. V. , Pauna E. I.
Published: December 2013

The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation.
Review and investigations of oscillatory flow behaviour of a horizontal ceiling opening for nuclear containment and fire safety analysis
Authors: Sharma P. K. , Singh R. K. , Ghosh A. K.
Published: December 2013

In the thermal hydraulics codes developed for fire safety analysis and for containment thermal hydraulic analysis, junctions in the multi-compartment geometries is often modeled as uni-directional junctions. However, ceiling junctions are known to depict unstable/oscillatory bi-directional flow behavior. Detailed investigations have been carried out to understand the unstable flow behaviour of a junction by analyzing an earlier reported experiment and its subsequent two dimensional numerical RANS based study of fire in an enclosure. The authors attempt more realistic and desired three dimensional and inherently transient large eddy simulations using a computer code Fire Dynamics Simulator (FDS). The paper presents the details of the analysis, the results obtained and further studies required to be conducted so that the findings can be applied to the fire/containment thermal hydraulics analysis codes successfully.
Neutron multiplication in source driven subcritical nuclear systems
Authors: Göksu A. , Geçkinli M.
Published: December 2013

In this work the neutron multiplication in a source driven subcritical nuclear system is studied for a particular physical model of symmetrical slab geometry. On both sides of a central spallation source region, bare fissile blankets are placed. The neutronics of the system is represented in terms of two-group diffusion equations. We elaborate on the development and evaluation of the multiplication factors for source and fission neutrons (and). The results are compared with the conventional multiplication factor of the fissile system. In the present static context, is important only from the safety point of view, whereas the other two factors are figures of merit measuring the effectiveness of a source neutron to start a fission chain and that of any one fission neutron to maintain a fading chain which further contributes to source multiplication.
Analysis of reactor lattice and core parameters in view of nuclear data modifications
Authors: Becker M.
Published: December 2013

A detailed assessment of the JEFF 3.1 and ENDF/B VII.0 nuclear data libraries with the KANEXT code system has been performed on each level of the reactor simulation steps.

This allowed for the sensitivity estimation of the lattice and core parameters to changes in nuclide and specific reaction cross section. The direct calculation of cross section data influence, in particular for specific reaction types, is possible due to the unique substitution feature of the group constant module GRUCAL.

Further on, the different impact of several reaction types on the lattice versus core calculation points out the “engineering uncertainty” using the two major libraries for the identical problem.

MCNP(X) and KANEXT cell models can be mutually adjusted such that differences between stochastic and deterministic data processing can be identified.

The current analysis emphasizes, that the two major nuclear data libraries JEFF 3.1 and ENDF/B VII.0 exhibit to some extent compensation of deviations rather than reliable integral data calculation.

On a special type of a fast reactor application we show that exchange of JEFF 3.1 and ENDF/B VII.0 cross sections introduce large discrepancies. A major part is attributed to the inelastic cross section. Further investigated discrepancies concerning the capture and the fission cross section in the epithermal energy range might be also a concern for LWR which was beyond the scope of the current study.
CANDU reactors with reactor grade plutonium/thorium carbide fuel
Authors: Şahin S. , Khan M. J. , Ahmed R.
Published: December 2013

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Validation of mechanistic CHF models using optical measuring techniques
Authors: Bloch G. , Sattelmayer T.
Published: December 2013

Critical heat flux (CHF) is an important effect in boiling heat transfer, but the effects causing it still remain to be fully understood. An overview of current mechanistic CHF models is presented, and experiments for validation are discussed.
Effect of gamma-radiation on sorption and precipitation of radionuclides
Authors: Kepák F.
Published: December 2013

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

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

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

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.
Phase separation in Si–(B)–C–N polymer-derived ceramics
Authors: Golczewski Jerzy A. , Aldinger Fritz
Published: December 2013

Details of phase separation in the microstructure of amorphous Si–(B)–C–N ceramics derived from polymers have been resolved using the recent results of structural investigations. The formation of an amorphous phase built of atomic compounds SiC
i/4N(4–i)/3 and consequently located along the composition line between SiC and Si3N4 in the ternary Si–C–N phase diagram demonstrates a generic feature of phase separation in all these materials. The amorphous carbon phase separates as a counterpart in the microstructure of Si–C–N ceramics, and in the case of Si–B–C–N ceramics such counterpart represents B–N–C domains of the composition (BN)
c
C
y
located along the tie line C–BN in the ternary B–C–N phase diagram. The effect of phase separation has been also pondered as a source of exceptional material properties.
Further results on creep behaviour of sand-cast Mg–2.8Nd–0.8Zn–0.5Zr–0.3Gd alloy at 0.56 to 0.61T
m under stresses 40 to 90MPa
Authors: Thomas-Whittington D. , Srivastava V. , Greenwood G. W. , Jones H.
Published: December 2013

An earlier study of the creep behaviour of Elektron 21 alloy has been extended to 290°C (0.61T
m; T
m: melting temperature). The combined results confirm a stress exponent of creep rate close to 6, but with an activation energy for creep of (310 ± 20) kJ/mol for 240 to 290°C and 40 to 90 MPa. Possible mechanisms giving rise to this behaviour are discussed.
Sigma phase formation and its effect on mechanical properties in the corrosion-resistant superalloy K44
Authors: Hou J. S. , Guo J. T. , Zhou L. Z. , Ye H. Q.
Published: December 2013

The sigma phase (σ) formation in long-term aged polycrystalline K44 superalloy has been studied as a function of aging temperature and time. The σ phase in K44 alloy forms mainly in dendrite cores with a few at interdendritic regions. The kinetics of σ formation can be described by the Johnson–Avrami–Mehl equation that yields activation energies of more than 130 kJ/mol in an early stage and less than 100 kJ/mol in steady state. Tensile experiments at room temperature and 1173 K and endurance experiments at 1173 K and 274 MPa were performed to test the effect of σ phases on these properties and no remarkable harmful effect is found. Moreover, the effect of long-term aging on the mechanical properties is realized mostly through the variation of γ′, not σ phase formation.
Structure-induced order–disorder transformation in Cd–Na liquid alloys
Authors: Singh R. N. , Ali I.
Published: December 2013

The origins of the S-shaped entropy of mixing and order–disorder transformation in Cd–Na liquid alloys have been investigated within a first-principle approach of a pseudopotential perturbation scheme with the Gibbs–Bogoliubov variation technique. We find that the deepening of the structure part of the entropy around the stoichiometric composition (c
Na ≍ 0.33) is responsible for the S-shaped entropy and suggests the existence of short-range order in the liquid state. The role of the local-field correction function, G(q), of conduction electrons on the entropy of mixing is examined. The impact of G(q) on partial structure factors is more visible towards low q(q → 0) than at higher q.
Comparison of thermodynamic data of the ternary Cu–Sn–Zn system, measured with the EMF and with the calorimetric method
Authors: Knott Sabine , Mikula Adolf
Published: December 2013

The enthalpy of mixing of the ternary Cu–Sn–Zn system was measured using a Calvet-type calorimeter. The investigations were carried out along the three cross sections Cu:Sn=1:2, 1:1, and 2:1. Zn was added in the concentration range between 0 and 50at.%.

The calorimetric results were compared with EMF data investigated in an earlier work by Peng and Mikula [1]. The results of the calorimetric measurements show very good agreement in the copper-rich region, compared with the EMF measurements, but in the Cu:Sn=1:2 cross section the results do not fit perfectly. The difference between both measuring methods was at the most ΔH = 300J/mol at the cross section Cu:Sn=1:1 and ΔH=150 J/mol at the Cu:Sn=2:1 cross section.
Site preference, thermodynamic, and magnetic properties of the ternary Laves phase Ti(Fe1–xAlx)2 with the crystal structure of the MgZn2-type
Authors: Yan Xinlin , Chen Xing-Qiu , Grytsiv A. , Witusiewicz V. T. , Rogl P. , Podloucky R. , Pomjakushin V. , Giester G.
Published: December 2013

Site preference and thermodynamic properties of the ternary Laves phase Ti(Fe1–xAlx)2 with MgZn2-type have been studied employing Rietveld refinement of X-ray and neutron powder diffraction data, X-ray single crystal data, and isoperibolic drop calorimetry techniques. A detailed relation between lattice parameters of Ti(Fe1–xAlx)2 and the Al content in the Laves phase has been presented. The Rietvelt refinement revealed that Ti atoms occupy the 4f site in the MgZn2-type, whilst Fe and Al atoms randomly share the 2a and 6h site. Heat of formation has been measured for Ti(Fe0.5Al0.5)2. For the ab initio density functional theory applications, a large number of structural models were investigated to calculate the concentration dependent (Al) heats of formation, magnetic structural stabilities, lattice parameters, and site occupancies, which are in good agreement with experiment.