Published by De Gruyter

Volume 80 Issue 3

Issue of Kerntechnik

Contents
Journal Overview

Contents/Inhalt

June 26, 2015

Page range: 187-189

Summaries/Kurzfassungen

June 26, 2015

Technical Contributions/Fachbeiträge

June 26, 2015

Improved data evaluation methodology for energy ranges with missing experimental data

A. Yu. Konobeyev, U. Fischer, R. Capote

Page range: 194-200

Abstract

A number of improvements of the data evaluation concerning nuclear model calculations and evaluation procedures are considered. A promising combination of the hybrid Monte Carlo simulation model of M. Blann for the modelling of non-equilibrium particle emission and the Hauser-Feshbach model, and a possible correction of simulations using intranuclear cascade evaporation model are discussed. In the last case the modelling of particle emission is improved by the consideration of nucleon-cluster interactions. The approach for reliable predictions of cross-sections using “optimal” nuclear model parameters and the method of their generation is discussed.

June 26, 2015

Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems

S. Dulla, Siew Sin Hoh, M. Nervo, P. Ravetto

Page range: 201-207

Abstract

The reactivity monitoring is a key aspect for the safe operation of nuclear reactors, especially for subcritical source-driven systems. Various methods are available for both off-line and on-line reactivity determination from direct measurements carried out on the reactor. Usually the methods are based on the inverse point kinetic model applied to signals from neutron detectors and results may be severely affected by spatial and spectral effects. Such effects need to be compensated and correction procedures have to be applied. In this work, a new approach is proposed, by using the full information from different local measurements to generate a global signal through a proper weighting of the signals provided by single neutron detectors. A weighting techique based on the use of the adjoint flux proves to be efficient in improving the prediction capability of inverse techniques. The idea is applied to the recently developed algorithm, named MAρTA, that can be used in both off-line and on-line modes.

June 26, 2015

Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant

A. M. Shokr, V. M. Carr

Page range: 208-213

Abstract

The feedback from the accident at the Fukushima-Daiichi nuclear power plant is crucial for defining and implementing measures for preventing accidents involving large releases of radioactive material at nuclear installations, including nuclear fuel cycle facilities. Following the lessons learned from this accident, assessment of the safety of nuclear fuel cycle facilities is essential to evaluate the robustness of the facilities' protection systems and components against the impact of extreme external events. A methodology to perform this safety assessment is presented, with discussions on possible preventive measures to be applied and mitigatory actions to be taken for further improvement of the robustness of nuclear fuel cycle facilities when subjected to extreme external events. Considerations in the assessment of multi-facility sites and use of a graded approach, commensurate with the facility's potential hazard, in application of the safety assessment methodology are also discussed.

June 26, 2015

The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant

H.-T. Lin, S.-M. Yang, J.-R. Wang, S.-W. Chen, C. Shih

Page range: 214-231

Abstract

In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1 088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1 100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

June 26, 2015

Development of a hydrogen diffusion gothic model of MARK III-containment

H. Zhen-Yu, H. Yu-Kai Huang, H. Wen-Sheng, Yen-Shu Chen, P. Bau-Shei

Page range: 232-240

Abstract

The accident that occurred at the Fukushima Daiichi Nuclear Power Plant is a reminder of the danger of hydrogen explosion within a reactor building. Sufficiently high hydrogen concentration may cause an explosion that could damage the structure, resulting in the release of radioisotopes into the environment. In the first part of this study, a gas diffusion experiment was performed, in which helium was used as the working fluid. An analytical model was also developed using the GOTHIC code and the model predictions of the helium distribution were found to be in good agreement with the experimentally measured data. In the second part of the study, a model of the Mark III containment of the Kuosheng Plant in Taiwan was developed, and was applied to a long-term station blackout (SBO) accident similar to that of the Fukushima plant. The hydrogen generation was calculated using the Modular Accident Analysis Program and was used as the boundary condition for the GOTHIC containment model. The simulation results revealed that the hydrogen concentration at the first floor of the wetwell in the containment reached 4% 9.7 h after the accident. This indicated the possibility of dangerous conditions inside the containment. Although active hydrogen ignitors are already installed in the Kuosheng plant, the findings of this study indicate that it may be necessary to add passive recombiners to prolong an SBO event.

June 26, 2015

Analysis of safety parameters for the MINERVE reactor

A. E. Ateya, E. H. Amin, M. E. Nagy, H. H. Abou-Gabal

Page range: 241-254

Abstract

The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO 2 , R1MOX, R2UO 2 , and R2MOX core configurations, representative of a LWR loaded with UO 2 , mixed oxide matrix, over moderated UO 2 and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.

June 26, 2015

Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378

P. U. Singare

Page range: 255-262

Abstract

The present paper demonstrates nondestructive application of the 131 I and 82 Br radioisotopes in performance evaluation of organic base anion exchange resins Purolite NRW-4000 and Duolite A-378. The evaluation was based on kinetics of iodide and bromide ion-isotopic exchange reactions taking place between the external ionic solution and the resin surface. It was observed that at a constant temperature of 35.0 °C when the concentration of bromide ions in solution increases from 0.001 mol/L to 0.004 mol/L, the specific reaction rate values for bromide ion-isotopic exchange increases from 0.228 to 0.266 min −1 using Purolite NRW-4000 resins and from 0.160 to 0.185 min −1 using Duolite A-378 resins. Also it was observed that under the same experimental conditions, the percentage of bromide ions exchanged increases from 60.67 % to 67.00 % using Purolite NRW-4000 resin, while for Duolite A-378 resin it increases from 49.30 % to 52.00%. The identical trend was observed for the two resins during iodide ion-isotopic exchange reaction. The overall results indicate that under identical experimental conditions, as compared to Duolite A-378 resins, Purolite NRW-4000 resins shows superior performance. It is expected that the present nondestructive radiotracer technique can be extended further for characterization of series of nuclear as well as non-nuclear grade ion exchange resins. The results of such study will be immensely useful in proper selection of the ion exchange resins in various industrial applications.

June 26, 2015

Neutronic performance of (^ReprocessedU/Th)O₂ fuel in CANDU 6 reactor

Z. Gholamzadeh, S. M. Mirvakili, S. A. H. Feghhi

Page range: 263-269

Abstract

Utilization of thorium-based fuel in different reactors has been under investigation for several decades. In fact, excellent breeding features, rather flattened distribution of power as well as proliferation resistance of such fuel cycle draws the attention towards utilization of this type of fuel in nuclear power technology. In the present study, the neutronic performance of a typical thorium core loading is addressed. In this configuration, a mixed uranium and thorium oxide is loaded in CANDU 6 reactor. The obtained results determine a total peaking factor of 2.73 for the proposed configuration. The values obtained for the β and the β eff are 332 and 303 pcm respectively. The core reactivity coefficients were more negative comparing the CANDU 6 loaded with nat UO 2 . The initial fissile material loaded in the core increased by a factor of 1.5 after 730 GWd burn-up. The obtained burn-up results show the core reactivity variations were highly positive after 6 and 12 h shut down because of considerably high buildup of 233 Pa after 1-year core operation at 2000 MW power.

June 26, 2015

Modelling study on production cross sections of ¹¹¹In radioisotopes used in nuclear medicine

A. Kara, T. Korkut, M. Yiğit, E. Tel

Page range: 270-274

Abstract

Radiopharmaceuticals are radioactive drugs used for diagnosis or treatment in a tracer quantity with no pharmacological action. The production of radiopharmaceuticals is carried out in the special research centers generally using by the cyclotron systems. Indium-111 is one of the most useful radioisotopes used in nuclear medicine. In this paper, we calculated the production cross sections of 111 In radioisotope via 111–114 Cd(p,xn) nuclear reactions up to 60 MeV energy. In the model calculations, ALICE/ASH, TALYS 1.6 and EMPIRE 3.2 Malta nuclear reaction code systems were used. The model calculation results were compared to the experimental literature data and TENDL-2014 (TALYS-based) data.

June 26, 2015

Study of ⁶⁰Co as gamma source in backscatter gamma densitometers

R. Gholipour Peyvandi, S. Z. Islami Rad, A. Taheri, S. Rahmanzadeh Tootkaleh, M. Askari Lehdarboni

Page range: 275-279

Abstract

In this work the performance of backscatter gamma densitometer is studied using 60 Co as gamma source. The ability of the densitometer to determine the liquid's interfaces in multi-phase flows was assessed. The aim mainly was to investigate the possibility of using 60 Co in this system. Furthermore, a comparison between 60 Co and was done. The obtained results indicate that 60 Co can be advantageous in this regard.

June 26, 2015

Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey

C. Eke, I. Boztosun

Page range: 280-290

Abstract

In this paper, we attempt to determine the activity concentration of natural and artificial radionuclides in 37 sand samples from the Mediterranean coast of Antalya in Turkey by using a high purity Germanium (HpGe) detector. 238 U, 232 Th, 40 K and 137 Cs activity concentrations, absorbed dose rate, annual effective dose equivalent, radium equivalent activity, external and internal hazard index of sand samples are determined res-pectively. The average values are 13.43 ± 0.21 Bq/kg, 6.96 ± 0.06 Bq/kg and, 122.46 ± 18.58 Bq/kg, for 238 U, 232 Th and 40 K respectively. Most of the activity concentration values are less than below minimum detection limit for 137 Cs. The average values of the absorbed dose rate D (nGy/h), annual effective dose equivalent AEDE(μSv/y), radium equivalent activity Ra eq (Bq/kg), external hazard index H ex and internal hazard index H in are 15.52 nGy/h, 19.03 μSv/y, 32.81 Bq/kg, 0.09 and 0.12 respectively. It is observed that 238 U, 232 Th, 40 K and 137 Cs activity concentrations, absorbed dose rate, annual effective dose equivalent, radium equivalent activity are in the limit of the published values, external and internal hazard index values are less than unity.

Technical Notes

June 26, 2015

Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions

S. A. Mousavi Shirazi

Page range: 291-294

Abstract

In this symbolic investigation, a cylindrical cell in a LWR, which consists of one fuel pin and moderator (water), is considered. The width of this cylindrical cell is divided into 100 equal units. Since the neutron flux in a cylindrical fuel pin is resulting from the diffusion equation: -1rddrφ(r)+∑aφ(r)=S(r)$ - \frac{1}{r}\frac{d}{{dr}}\varphi \left( r \right) + \mathop \sum \nolimits_a \varphi \left( r \right) = S\left( r \right)$, the amount of fast neutron fluxes are obtained on the basis of the numeric solution of this equation, and the applied boundary conditions are considered: φ′(0)=φ′(1)=0$\varphi '\left( 0 \right) = \varphi '\left( 1 \right) = 0$. This differential equation is solved by the tridiagonal method for variant enrichments of uranium. Neutron fluxes are obtained in variant radii of fuel pin and moderator and are finally compared with each other. There are some interesting outcomes resulting from this investigation. It can be inferred that because of the fuel enrichment increment, the fast neutron flux increases significantly at the centre of core, while many of the fast neutrons produced are absorbed after entering the water region, moderation of lots of them causes the reduced neutron flux to get improved in this region.

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.

Volume 80 Issue 3

Issue of Kerntechnik

Contents

Summaries

Improved data evaluation methodology for energy ranges with missing experimental data

Abstract

Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems

Abstract

Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant

Abstract

The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant

Abstract

Development of a hydrogen diffusion gothic model of MARK III-containment

Abstract

Analysis of safety parameters for the MINERVE reactor

Abstract

Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378

Abstract

Neutronic performance of (ReprocessedU/Th)O2 fuel in CANDU 6 reactor

Abstract

Modelling study on production cross sections of 111In radioisotopes used in nuclear medicine

Abstract

Study of 60Co as gamma source in backscatter gamma densitometers

Abstract

Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey

Abstract

Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions

Abstract

Neutronic performance of (^ReprocessedU/Th)O₂ fuel in CANDU 6 reactor

Modelling study on production cross sections of ¹¹¹In radioisotopes used in nuclear medicine

Study of ⁶⁰Co as gamma source in backscatter gamma densitometers