Published since January 1, 1987

Kerntechnik

Independent Journal for Nuclear Engineering

ISSN: 2195-8580

Impact Factor: 0.293

About this journal

Kerntechnik is an independent journal for nuclear engineering (including design, operation, safety and economics of nuclear power stations, research reactors and simulators), energy systems, radiation (ionizing radiation in industry, medicine and research) and radiological protection (biological effects of ionizing radiation, the system of protection for occupational, medical and public exposures, the assessment of doses, operational protection and safety programs, management of radioactive wastes, decommissioning and regulatory requirements). For more than 75 years Kerntechnik offers original scientific and technical contributions, review papers and conference reports.
All articles are subject to thorough, independent peer review.

History

In 1956, the publisher Karl Thiemig founded the journal ATOMKERNENERGIE after the Paris Treaties of May 1955 enabled the Federal Republic of Germany to re-enter nuclear technology. The dynamic scientific development that started at that time prompted Karl Thiemig to launch a second technical journal, KERNTECHNIK. At the end of 1970, the journal ATOMPRAXIS, published by Verlag Braun, was merged into KERNTECHNIK.

In 1979, the journals ATOMKERNENERGIE and KERNTECHNIK in turn merged into one journal with the ATOMKERNENERGIE/KERNTECHNIK. When the Thiemig publishing house ceased its operations in early 1986, Carl Hanser Verlag acquired the journal and continued to publish it under the title KERNTECHNIK from 1987.

KERNTECHNIK is the only independent German journal publishing scientific articles in English in the fields of nuclear technology and radiation protection. During the last decade, KERNTECHNIK has increasingly developed into a journal respected by international experts. In order to secure KERNTECHNIK for the future, in which high visibility of publications, the impact factor and open access play an increasingly decisive role, Carl Hanser Verlag has decided to place the journal KERNTECHNIK into the hands of the very renowned and successful scientific publisher Walter de Gruyter as of 01.01.2021.

De Gruyter publishes first-class research and has done so for more than 270 years and is an international, independent publisher headquartered in Berlin. The company publishes over 1,300 new book titles each year and more than 900 journals in the humanities, social sciences, medicine, mathematics, engineering, computer sciences, natural sciences, and law, and also offers a wide range of digital media, including open access journals and books.

The publishing house is represented in international markets by its own offices or partners. As an independent scientific publisher with a long-term strategy, de Gruyter offers an almost perfect environment for KERNTECHNIK to flourish and is an excellent strategic addition to its technology and natural sciences division.

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Topics

Volume 86 Issue 4

August 2021

Accessible August 18, 2021

Frontmatter

Page range: 257-322

Download PDF

August 18, 2021

Large eddy fire simulation applications from nuclear industry

P. K. Sharma, V. Verma, J. Chattopadhyay, G. Vinod

Page range: 260-272

Abstract

A computational study has been carried out for predicting the behaviour of a pool fire source using the field-model based code Fire Dynamics Simulator (FDS). Time dependent velocity and temperature fields are predicted along with the resulting changes in the plume structure and its width. Firstly, a grid study was performed to find out the best grid size for this purpose. Then calculations were done which showed a very good agreement with earlier reported experimental based correlations for the temperature of the plume region. These studies have been extended to use this field-model based tools for modelling particular separate effect phenomena like puffing frequency and to validate against experimental data. There are several applications in nuclear industry like room fires, wildland fires, smoke or ash disposal, hydrogen transport in nuclear reactor containment, natural convection in building flows etc. In this paper the use of FDS with the advanced Large Eddy Simulation (LES) based CFD turbulence model is described for various applications: Fire simulation for Alpha storage, Bhabhatran teletherapy, pool fire for transport casks, fire PSA of a representative NPP, exhaust air fan buildings of a process plant and smoke dispersion in large fires around NPPs.

August 18, 2021

Characterization of polystyrene and polyacrylic based polymeric materials exposed to oxidative degradation

P. U. Singare

Page range: 273-282

Abstract

The characterization of oxidative degraded polystyrene-based resin (R1) and polyacrylic based resin (R2) resins in H 2 O 2 and HClO 4 degradation medium were made based on the kinetics and thermodynamic data obtained for the ion-isotopic exchange reactions using such resins. For the reactions performed by using resins degraded in H 2 O 2 medium, the reaction rate (k) values obtained for the fresh R1 (0.315 min –1 ) and R2 (0.187 min –1 ) resins decreases to 0.300 and 0.155 min –1 respectively for the resins degraded in 20% H 2 O 2 medium, which further decreases to 0.289 and 0.142 min –1 respectively for the resins degraded in 30% H 2 O 2 medium. A similar trend in the results were observed for the reactions performed by using the above resins degraded in HClO 4 medium. The higher values of k (min –1 ) and low values of various thermodynamic parameters for the ion-isotopic exchange reactions performed by using fresh and degraded polystyrene-based resin R1 resins suggests superior degradation stability as compared to polyacrylic based R2 resin.

August 18, 2021

Thorium-based CANDU qualification as plutonium burner

R. Neacşa, A. Rizoiu, I. Prisecaru

Page range: 283-293

Abstract

Converting the weapon grade Plutonium from the U.S.A., Russia, U.K. etc. to M ixed OX ide fuel and using it in power reactors was seen as a feasible way to both dispose Plutonium and produce energy. Using Thorium-based fuels in CANDU has been investigated since early 1980’s, they were designed and tested in Canada as mixed ThO 2 -UO 2 (both LEU and HEU) and mixed ThO 2 -PuO 2 , (both reactor- and weapons-grade) ([1]). In this respect, Thorium might also be seen as a valuable driver for weapon grade Plutonium annihilation. Our goal was to investigate ThO 2 -PuO 2 MOX in the aim to propose a suitable fuel for the existing and future CANDU units in Romania. Both weapon grade and reactor grade Plutonium were considered as fissile drivers for Thorium. Since this is only an exploratory study, some key design parameters such as fuel pellet density and ThO 2 /PuO 2 ratio were considered to span over a certain range imposed by MOX fuel fabrication technology and limited Plutonium availability. Eighteen fuel compositions were considered and cell calculations were performed for 37 and 43-element bundles using several computer codes.

August 18, 2021

Higher order T_N approximation for the neutron diffusion problem in a slab reactor

H. Öztürk, B. Durmaz

Page range: 294-301

Abstract

Higher order approximations of the Chebyshev polynomials of first kind (T N ) are used for the first time in calculation of the diffusion lengths of monoenergetic neutrons in a homogeneous slab. In the method, the diffusion lengths of the neutrons are calculated using various values of the c, the number of secondary neutrons per collision. First, the traditional Legendre polynomials (P N ) approximation and then the present T N method are used separately. The numerical results for the diffusion lengths are tabulated in the tables up to an order of N = 9. A brief comparison is also done between the results obtained from the present method and the ones in literature. The advantages of the present method can easily be observed from the good accordance between results given in the tables for comparison and its easily executable equations. For many of the c values, the results obtained from T N method are better than the results obtained from P N method.

August 18, 2021

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

M. E. Korkmaz, N. K. Arslan

Page range: 302-311

Abstract

Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MW th total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

August 18, 2021

Neutronics and both analytical and numerical solutions for the rod centered subchannel thermal-hydraulic model

M. Y. M. Mohsen, M. A. E. Abdel-Rahman

Page range: 312-320

Abstract

This work investigates the capacity to detect safety levels in pressurized water reactors using neutronic and thermal-hydraulic calculations (PWR). The main neutronic parameters, such as criticality calculations, reactivity calculations, burn-up calculations, and power distribution calculations, have been calculated by using MCNPX based on the Monte Carlo method. The hot channel was determined from the power distribution. On the hot channel, the rod-centered sub-channel model was used to investigate the main thermal-hydraulic parameters such as the fuel, cladding, and coolant temperature distribution in both axial and radial directions, the coolant pressure drop, and the departure from nucleate boiling ratio (DNBR). Two techniques were used to solve the rod-centered sub-channel model. MATLAB code was used for the analytical methodology, and COMSOL-Multiphysics computer software was used for the numerical methodology. The numerical technique has demonstrated high accuracy in determining the reactor safety level consistent with the analytical approach.

Volume 86 (2021)

Volume 85 (2020)

Volume 84 (2019)

Volume 83 (2018)

Volume 82 (2017)

Volume 81 (2016)

Volume 80 (2015)

Volume 79 (2014)

Volume 78 (2013)

Volume 77 (2012)

Volume 76 (2011)

Volume 75 (2010)

Volume 74 (2009)

Volume 73 (2008)

Volume 72 (2007)

Volume 71 (2006)

Volume 70 (2005)

Volume 69 (2004)

Volume 65 (2000)

Volume 64 (1999)

Volume 62 (1996)

Volume 61 (1996)

Volume 60 (1995)

Volume 59 (1994)

Volume 58 (1993)

Volume 57 (1992)

Volume 56 (1991)

Volume 55 (1990)

Volume 54 (1989)

Volume 53 (1989)

Volume 52 (1988)

Volume 51 (1987)

Volume 50 (1987)

5-year Impact Factor	0.234
Cite Score	0.6
Impact Factor	0.293
Journal Citation Indicator	0.17
SCImago Journal Rank	0.242
Source Normalized Impact per Paper	0.367

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Manuscript submission

You can easily submit your manuscript online. Simply go to https://mc.manuscriptcentral.com/kern and you will be guided through the whole submission and publishing process.

Editorial

Editor-in-Chief
Dr.-Ing. Anne Krüssenberg (Kerntechnik@degruyter.com)

(Deutsch)

Online ISSN: 2195-8580
Print ISSN: 0932-3902
Type: Journal
Language: English
Publisher: De Gruyter
First published: January 1, 1987
Publication Frequency: 6 Issues per Year

Kerntechnik

Independent Journal for Nuclear Engineering

Frontmatter

Large eddy fire simulation applications from nuclear industry

Abstract

Characterization of polystyrene and polyacrylic based polymeric materials exposed to oxidative degradation

Abstract

Thorium-based CANDU qualification as plutonium burner

Abstract

Higher order TN approximation for the neutron diffusion problem in a slab reactor

Abstract

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Abstract

Neutronics and both analytical and numerical solutions for the rod centered subchannel thermal-hydraulic model

Abstract

Higher order T_N approximation for the neutron diffusion problem in a slab reactor