bluuebunny/crossref_metadata_2025_split huggingface dataset:
CANDU fuel elements behaviour in the load following tests
Authors: Horhoianu G. , Palleck S. 
Published: December 2013

Abstract
Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during test period. New LF tests are planed to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions.
A study on the damage of potential first wall materials in a nuclear fusion reactor using plutonium bearing salt
Authors: Übeyli M. 
Published: December 2013

Abstract
Selection of first wall material for fusion reactors is very crucial when taking into account of fusion blanket design and operation cost. A realistic way to extend the working period of first wall structure is to use a protective flowing liquid wall between fusion plasma and first wall. HYLIFE-II, one of the important fusion reactor design concepts, uses such a liquid wall. In the current article, the radiation damage on the first wall of HYLIFE-II fusion reactor was investigated for various candidate materials. In the liquid wall of the reactor, a molten salt containing weapon grade (WG) plutonium isotopes was used. The numerical results indicated that a refractory alloy of W-5Re was found to have the lowest damage values. In addition, the use of WG plutonium isotopes did not have a negative effect on the radiation damage characteristics of the investigated structural materials.
Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment
Authors: Thangamani I. , Gera B. , Dutta A. , Verma V. , Singh R. K. , Ghosh A. K. 
Published: December 2013

Abstract
The proposed Advanced Heavy Water Reactor (AHWR) employs double containment envelope along with many Engineered Safety Features (ESFs) to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) with safety system failure, during which high enthalpy steam and radioactive fission products will be discharged into the containment. In such conditions, the pressurized containment will be the source of activity release to the environment by way of leakage. It is required to study the effect of ESFs on the source term from the AHWR containment. An analysis was performed to evaluate the release rate from the AHWR containment during a postulated accident with the in-house containment code CONTRAN and the aerosol behavior code NAUA5-M in a coupled way. Modules for simulating the engineered safety features were incorporated in the CONTRAN code and the aerosol transport behaviour was evaluated using NAUA5-M separately. The AHWR containment is divided into three nodal volumes interconnected by junctions. The blow down mass, energy discharge data and activity released into the containment from the reactor core, for a postulated LOCA case of 200% RIH break with failure of shutdown systems (1 & 2), are inputs to the CONTRAN code. Thermodynamic parameters like containment gas temperature, partial pressure of steam, air in the subdivided volumes along with the flow rates through junctions obtained from CONTRAN were supplied to NAUA5-M. An analysis was carried out for a number of cases, postulated based on availability/unavailability of ESFs. Pressure, temperature and activity concentration transients were evaluated, for 72 h, in the subdivided volumes along with the activity released out of the containment through leakages and stack discharges for all the cases. This paper highlights the importance of operation of ESF in reducing the activity release to the environment.
Determination of 89Zr production parameters via different reactions using ALICE and TALYS codes
Authors: Taghilo M. , Kakavand T. , Sadeghi M. 
Published: December 2013

Abstract

89Zr is an important positron-emitting radionuclide for positron emission tomography (PET) and used in the field of tumor diagnostics, tumor therapy and the investigation of the bio-kinetic. The ALICE-91 and TALYS-1.0 codes were used to calculate excitation function for proton, alpha, deuteron and neutron induced on various targets that lead to the production of 89Zr radioisotopes using intermediate energy accelerators. Requisite thickness of the targets was obtained by SRIM code. The 89Zr production yield was evaluated using excitation function and stopping power. For the 89Y(p, n)89Zr reaction, the calculation data were compared with the experimental data. The 89Y(p, n)89Zr process was determined as most interesting one due to radionuclide purity. The ALICE-91 and TALYS-1.0 codes predict a maximum cross-section of about 939.86 mb at 15MeV and 860 mb at 13MeV respectively for this reaction.
Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading
Authors: Pabarcius R. , Tonkunas A. , Slavickas A. , Seporaitis M. 
Published: December 2013

Abstract
Starting from early to mid 1970s the first generation of channel-type graphite-moderated boiling water reactors socalled RBMK types were designed and went into operation. Nowadays, all designed RBMKs represent three generations of reactors having significant differences with respect to their safety design features. The Ignalina NPP with two RBMK-1500 units being a second generation type were originally the most powerful reactors in the world. The change of important neutron-physical characteristics, being specific safety features for RBMK type reactors during RBMK-1500 core upgrading processes are analyzed until final shutdown and their impact on safety during reactor lifetime is discussed.
Training of the KBR nuclear emergency staff
Authors: Wink R. , Fary R. 
Published: December 2013

Abstract
In order to respond to a nuclear emergency case the first few hours are of decisive significance. For that reason sufficient and well-trained staff must be available at any time. The training of the personnel of the Nuclear power plant Brokdorf (KBR) involved in the emergency management organisation will be described in this presentation.
Preparation of emergency care centre exercises
Authors: Schnadt H. , Miska H. 
Published: December 2013

Abstract
Setup and operation of emergency care centres (sometimes also addressed as emergency reception centres) are part of emergency response in the environs of nuclear power plants. The preparation of an exercise scenario for such a centre is very demanding on the responsible agency. Therefore, a computer code has been developed which helps to translate the exercise objectives into instructions for figurants which simulate the affected population. These instructions are intended to steer a determined flow of people through the emergency care centre by providing fictitious radiological readings and by injecting the demand for additional actions of response personnel by statements and questions.
Independent assessor's report on the experience with preparation, managing, observing and evaluation of emergency exercises without prior notice in nuclear power plants
Authors: Matzig B. , Solisch T.-O. 
Published: December 2013

Abstract
We inspect regularly the measures taken in terms of the plants operator emergency management organisation in supposed emergency situations and assess their effectiveness on behalf of the supervisory authorities under the atomic energy law. To this end corresponding exercises based on scenarios that have been prepared by ESN are carried out in the plants and followed by a group of independent observers. The course and effectiveness of measures initiated are assessed in the form of an expert's report. ESN prepared, managed, observed und evaluated more than 75 emergency exercises in nuclear power plants in the last twenty years. With this experience and particularly with the experience feedback between plants we contributed to the development of the internal emergency management for nuclear power plants in Germany. This paper at hand summarizes some of our experience.
Planning, implementation and evaluation of an emergency exercise at the Paul Scherrer Institute
Authors: Hirt R. , Lörtscher Y. 
Published: December 2013

Abstract
Like all nuclear power plants in Switzerland, the Paul Scherrer Institute in Villigen (PSI), Switzerland, has the duty to conduct an emergency exercise every year. The two monitoring authorities ENSI (Swiss Federal Nuclear Safety Inspectorate) and the BAG (Federal Office of Public Health – FOPH) use this exercise for their yearly audit of the emergency organization. The emergency organization of the Paul Scherrer Institute consists of the emergency staff, the fire brigade, the first aid station, the traffic- and the security platoon and the radiation safety, the latter consisting of the decontamination group, the radio analytics group and the measurement group (for radioactivity dispersion calculations). The framework is set up by the ENSI beforehand, based on a fixed cycle. For the emergency exercise of 2009 a plane crashing into the interim storage facility of the federal government (BZL) at PSI was assumed. This contribution describes the approach of the planning process, the performance and the evaluation of the exercise.
Simulation of HALDEN IFA-650 loss-of-coolant accidents tests with TRACE
Authors: Aounallah Y. 
Published: December 2013

Abstract
The trends towards higher burnups, longer in-core fuel residence time, and the development of new cladding materials, have led to the need to re-assess the safety criteria for loss-of-coolant accidents. The HALDEN IFA-650 series of experiments are integral, single-pin, in-pile tests on the fuel behavior undergoing conditions of a loss-of-coolant accident over pre-determined fuel rod linear heat rate, and under thermally-controlled structural environment. Fresh and irradiated fuel pins from commercial reactors were tested. The present work seeks to assess the capability of a new (best-estimate) thermal-hydraulics code TRACE (version 5.0), to adequately predict the thermal-hydraulic impact of a loss-of-coolant accident on fuel rods. Several tests have been simulated over a range of fuel rod power and surrounding structure temperatures. The tests initiated from conditions of single-phase liquid natural circulation operating at 7MPa with subsequent venting to near-atmospheric pressure.
Upgrading (V)HTR fuel elements for generationIV goals by SiC encapsulation
Authors: Knorr J. , Kerber A. , Moormann R. 
Published: December 2013

Abstract
The pebble bed reactor is one of the most promising concepts for (V)HTR. Nevertheless recent re-evaluation of AVR and THTR results emphasizes once more that claimed advantages strongly depend on the properties of the fuel elements and their behavior under operational conditions. Additionally safeguards, waste management and disposal aspects gain increasing importance today. The conventional uncoated graphite pebbles meet only inadequately the requirements of Generation IV facilities. Since long experts agree, that corrosion-resistant pebbles with high retention capability for fission products would considerably improve the chances of the pebble bed reactor concept. With laser beam joining of ceramics the key technology is now available for the silicon carbide (SiC) encapsulation of (V)HTR components. The envisaged innovative fuel element consists of a robust SiC hollow sphere filled with moderator and TRISO coated particles. The positive assessment according to Gen IV criteria should justify necessary R&D efforts to obtain qualified fuel elements and demonstrate their superiority under operational conditions.
Slug modeling with 1D two-fluid model
Authors: Tiselj I. , Martin C. Samuel 
Published: December 2013

Abstract
Simulations of condensation-induced water hammer with one-dimensional two-fluid model requires explicit modeling of slug formation, slug propagation, and in some cases slug decay. Stratified flow correlations that are more or less well known in 1D two-fluid models, are crucial for accurate description of the initial phase of the slug formation and slug propagation. Slug formation means transition to other flow regime that requires different set of correlations. To use such two-fluid model for condensation induced water hammer simulations, a single slug must be explicitly recognized and captured. In the present work two cases of condensation-induced water hammer simulations performed with WAHA code, are described and discussed: injection of cold liquid into horizontal pipe filled with steam and injection of hot steam into horizontal pipe partially filled with cold liquid.
A discussion of hyperbolicity in CATHENA 4: Virtual Mass and phase-to-interface pressure differences
Authors: Aydemir N. U. 
Published: December 2013

Abstract
It is well known that the one-dimensional equations of motion for two-phase flow are non-hyperbolic. Non-hyperbolicity can lead to numerical instabilities, destroying the solution. However, researchers in the last few decades were able to show that inclusion of virtual mass and/or phase-to-interface pressure differences in the momentum equations successfully render the equations of motion hyperbolic. In the present paper, the effect of including virtual mass and phase-to-interface pressure terms in the momentum equations on the hyperbolicity of the two-phase model in the CATHENA 4 code is discussed. The study is motivated by the fact that the inclusion of either model has been shown in the open literature to lead to a hyperbolic system separately. However, no known study exists that examine hyperbolicity in the presence of both these terms in the momentum equations. In this work, both terms are considered in the model equations simultaneously and their implications on the hyperbolicity of the two-phase model are discussed. Specifically, it is shown that in the case of mixed flow, there is a distinct region of non-hyperbolicity that developers need to be aware of when their equations include both the virtual mass and the phase-to-interface terms. Selecting the coefficients of phase-to-interface pressure difference terms properly ensures that the equations are hyperbolic for a wide range of conditions.
Condensation induced water hammer – overview and own experiments
Authors: Dirndorfer S. , Doerfler M. , Kulisch H. , Malcherek A. 
Published: December 2013

Abstract
A condensation induced water hammer is a severe effect that has caused damages and even fatalities worldwide. This effect is not limited to the nuclear industry only. Generally it can occur in every pipe that contains water and steam (two-phase-flow). To date many experiments were executed to study this phenomenon and to validate existing water hammer codes. But theoretically versus measured results show a very high degree of deviation. Information about those experiments was obtained by literature review. This information is presented additionally to a test facility, which has been developed in order to generate water hammers. The test device is introduced in this paper and its main functions are briefly explained.
In contrast to other experiments, water hammers at the UniBw are performed under controlled conditions. Air in water is a disruptive parameter that causes vigorous deviations between the theoretical system code and the experimental measurements. As in the test device presented in this paper, deionised water without air is used, this effect is ruled out. Furthermore the inception of a water hammer is controlled by a scientist. The experimental results will characterise the dynamic response of the test device as a function of system pressure, filling degree and sub-cooling.
Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices
Authors: Hegyi G. , Keresztúri A. , Tóta A. 
Published: December 2013

Abstract
The experiments performed at the ZR-6 zero power critical reactor by the Temporary International Collective (TIC) and a burnup benchmark specified for depletion calculation of a VVER-440 assembly containing Gd burnable poison were used to qualify the APOLLO2.8-3.E (APOLLO2) code as a part of its ongoing validation activity. The work is part of the NURISP project, where KFKI AEKI undertook to develop and qualify some calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell, macro-cell and 2D calculations of selected regular and perturbed experiments are used for the validation. In the 2D cases, the radial leakage is also taken into account by the axial leakage represented by the measured axial buckling. Criticality parameter and reaction rate comparisons are presented. Although various sets of the experiments have been selected for the validation, good agreement of the measured and calculated parameters could be found by using the various options offered by APOLLO2. An additional mathematical benchmark – presented in the paper – also attests for the reliability of APOLLO2. All the test results prove the reliability of APOLLO2 for VVER core calculations.
The simplified P3 approach on a trigonal geometry of the nodal reactor code DYN3D
Authors: Duerigen S. , Fridman E. 
Published: December 2013

Abstract
DYN3D is a three-dimensional nodal code for steady-state and transient analyses of Light Water Reactors applicable for square and hexagonal fuel assembly geometries. Several versions of the DYN3D code are available including a multigroup diffusion and a simplified P3 (SP3) neutron transport option. The multi-group SP3 method based on a trigonal geometry was developed recently. This method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement. In this paper, the theoretical background for the SP3 method is briefly described. The consistency of the implementation of the trigonal SP3 methodology in DYN3D is demonstrated by means of a simplified VVER-440 core test example. The corresponding few-group homogenized cross sections and reference solutions were produced by the Monte Carlo code Serpent. The DYN3D results are in good agreement with the reference solutions.
Analysis of coolant flow in central tube of VVER-440 fuel assemblies
Authors: Zsíros G. , Tóth S. , Aszódi A. 
Published: December 2013

Abstract
A CFD model has been developed to investigate the coolant flow in the central tube of VVER-440 fuel assemblies. The model has been validated with measured data of Kurchatov Institute. With this model a peripheral and an inner fuel assembly of the core have been investigated and the mass flux and heatup ratios of the central tube flow have been determined. Based on this study, the outlet mass flux of the tube is 1.2 times higher than the inlet mass flux of the rod bundle and heat-up in the tube is 0.35 times lower than the heat-up in the rod bundle. The ratios are not sensitive to the operational conditions within the scope of these investigations. The results of these simulations can be used as boundary conditions for the central tube in the assembly head calculations.
Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC
Authors: Tusheva P. , Schäfer F. , Reinke N. , Altstadt E. , Kliem S. 
Published: December 2013

Abstract
The research field focussing on the investigations and the analyses of postulated severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises of the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. This paper is focused on the possibilities for accident management measures in case of severe accidents. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accidents, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed.
VVER-440 with inert matrix fuel – viable direction to sustainability
Authors: Dařílek P. , Strmenský C. , Zajac R. , Majerčík J. 
Published: December 2013

Abstract
Reutilisation of fissile nuclides after reprocessing in light water reactor VVER-440 with inert matrix fuel (IMF) is analysed. Development of transmutation combined fuel assembly, containing two pin types – with metal molybdenum + PuO2 mixture and with classical UO2 – is described. Static characteristics of five-year VVER-440 fuel cycle with IMF are given. Results of control rod ejection accident analysis are presented. Necessary improvements of transmutation combined fuel assembly with IMF for the reactor VVER-440 are mentioned.
The world we really live in – communication on radiation
Authors: Gellermann R. 
Published: December 2013

Abstract
In order to make the communication on radiation and radiation levels more tangible and clear to the general public, measured results from actual living environments are needed. The paper demonstrates by means of measured results of the ambient dose rates (ODL) that the ambient dose rates in the centre of German cities locally exceed levels typical for the respective geogenic environment significantly. The highest dose rates which have been measured up to now reach 680nSv/h. The role of these results from the radiation protection point of view is discussed and a level of 300nSv/h is suggested as an upper level of the normal ambient dose rate in the urban environment. Still higher dose rates have been found in areas with radioactively contaminated soil. The need for remediation of such contaminated sites is indicated.
Appraisal of long-term radiation trends in the environs of nuclear power plants – Examples from Switzerland
Authors: Bucher B. , Rybach L. , Schwarz G. 
Published: December 2013

Abstract
Annually since 1989, biannually since 1994 the sites of the Swiss nuclear facilities (4 power plant sites, a nuclear research institute and a waste storage facility) are surveyed flying the same survey lines by airborne gamma ray spectrometry. The equipment and the data processing software used for those surveys was built and developed at the Institute of Geophysics, ETH Zurich. For mapping the ground radiation around the nuclear facilities a pixel representation and a modified Spectrum Dose Index (SDI) method is used. In the search for long-term trends the local dose-rates are calculated first and in turn the net dose rates over the time span 1992–2010. So far no significant change in the radiation levels was detected over the last 18 years outside of the fenced sites of the nuclear facilities, except for the operation of an industrial radiation facility in 1995.