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Svadlenkova M.,Research Center Rez Ltd. | Heraltova L.,Czech Technical University | Juricek V.,Research Center Rez Ltd. | Kostal M.,Research Center Rez Ltd. | Novak E.,Research Center Rez Ltd.
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment | Year: 2014

A method of measurement and analysis of gamma spectra of short living fission products of lightly irradiated fuel pins has been developed at facilities at the LR-0 zero-power experimental reactor. Experimental and computational methods of the peak area correction for radioactive decay are described. Selection of energy peaks suitable for deriving power distribution in the core was performed. © 2013 Elsevier B.V.


Kucera J.,Research Center Rez Ltd | Kofronova K.,Police of the Czech Republic
Journal of Radioanalytical and Nuclear Chemistry | Year: 2011

Autopsy of 29-year old woman suspicious of committing suicide by the ingestion of As2O3 yielded contradictory findings. All pathological findings as well as clinical symptoms suggested acute poisoning, while a highly elevated As level of 26.4 μg g-1 in her hair collected at the autopsy, which was determined with inductively coupled plasma mass spectrometry indicated chronic poisoning. To elucidate this discrepancy, instrumental neutron activation analysis (INAA) with proven accuracy was performed of another set of sectioned hair samples. Levels of As found by INAA in the range of 0.16-0.26 μg g-1 excluded chronic poisoning, because the person died after approximately 14 h after the As2O 3 ingestion. Two reasons for the discordant As results obtained by ICP-MS and INAA are considered: (1) accidental, non-removed contamination of hair on the As2O3 ingestion; (2) erroneous performance of ICP-MS. © 2010 Akadémiai Kiadó, Budapest, Hungary.


Kost'Al M.,Research Center Rez Ltd. | Cvachovec F.,University of Defence at Brno | Rypar V.,Research Center Rez Ltd. | Juricek V.,Research Center Rez Ltd.
Annals of Nuclear Energy | Year: 2012

The neutron fluence load on a reactor pressure vessel is an important physical quantity affecting material degradation and consequently reliable assessment of pressure vessel integrity and lifetime prolongation beyond designed conditions. This degradation is influenced by mixed neutron-photon source from the core periphery as well as by material parameters of the reactor pressure vessel. Computational procedures and experimental determination of neutron fluxes in the VVER-1000 mock-up internal structures as well as the results achieved are described in this paper. The calculation were performed with the MCNPX code with different nuclear data libraries. Nuclear data were processed using NJOY code. The neutron spectra measurements were performed with a two-parameter stilbene spectrometer. © 2011 Elsevier Ltd. All rights reserved.


Kostal M.,Research Center Rez Ltd | Kostal M.,Nuclear Research Institute Řež | Rypar V.,Research Center Rez Ltd | Svadlenkova M.,Research Center Rez Ltd
Nuclear Engineering and Design | Year: 2012

The pin power density distribution in reactor is an important quantity, necessary for the adequate assessment of fuel conditions and of core structures and pressure vessel radiation embrittlement as well. The paper shows the detailed comparison of calculated and experimentally determined pin by pin power distribution. To verify the reliability of measured data used for comparison with calculated data, the symmetrically located pins were measured. The calculations have been done with deterministic and Monte Carlo approach. The effect of different data libraries used for calculations are discussed as well. © 2011 Elsevier B.V. All rights reserved.


Kostal M.,Research Center Rez Ltd. | Svadlenkova M.,Research Center Rez Ltd. | Milcak J.,Research Center Rez Ltd.
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment | Year: 2014

97Zr is a relatively high-yield fission product that can be used for zero reactor power determination. The technique is not widely used because in the case of reactors that use zirconium metal in the fuel cladding, it is not only a fission product but is also produced by activation. In an appropriately chosen time interval, results obtained using 97Zr can be compared to those of power determination performed using 92Sr. The knowledge of the ratio between fission-induced 97Zr and the portion of 97Zr activated in the cladding can be used not only for power-density determination but also as an important indication of fuel failures. © 2013 Elsevier B.V.


Jansky B.,Research Center Rez Ltd. | Novak E.,Research Center Rez Ltd.
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment | Year: 2014

Spherical hydrogen proportional detectors (HPD) of 40 mm diameter with different pressures are used in neutron spectrometry in energy range 20-1200 keV. This spherical shape detector having anode wire along one diameter was designed to give a response to neutrons which is independent to their direction of travel with respect to the counter axis (anode). Nevertheless, in practice some imperfectness of spherical detector response exists. The reactor iron filtered beam was assembled with the aim to study detector response dependent on the angle (90 and 0; perpendicular and parallel) between neutron beam and detector anode. It was confirmed that the dominant effect is caused by neutron shielding properties of the massive end-parts of detector (insulators, aluminum ending). This effect produces 7-13% decrease of registered neutrons depending on energy interval. The resonance energy structure with characteristic maxima of iron filtered beam is used for energy calibration. Comparison with He-3 admixture calibration method is presented. Negative energy shift of measured maxima in neutron beam spectrum is observed. Energy shift depends on high voltage, on the angle between neutron and detector anode and on energy of neutron peak. This shift effect is caused by gas multiplication saturation close to anode wire. © 2013 Elsevier B.V.


Kostal M.,Research Center Rez Ltd. | Rypar V.,Research Center Rez Ltd. | Juricek V.,Research Center Rez Ltd.
Annals of Nuclear Energy | Year: 2013

Adequate and accurate determination of multiplication coefficient is not a simple task. In the case of fuel facilities or reactor core designs as well as in the case of emergency shut-down system calculations keff overestimation may lead to an unduly conservative approach. In this paper the critical levels for VVER-1000 mock-up, assembled on zero-power reactor LR-0, with various concentrations of boric acid are presented. These experimental data are compared with calculations in different nuclear data libraries. © 2013 Elsevier Ltd. All rights reserved.


Kost'al M.,Research Center Rez Ltd. | Svadlenkova M.,Research Center Rez Ltd. | Milcak J.,Research Center Rez Ltd.
Applied Radiation and Isotopes | Year: 2013

The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5. W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density. © 2013 Elsevier Ltd.


Zhu X.-K.,Research Center Rez Ltd. | Leis B.N.,Research Center Rez Ltd.
International Journal of Pressure Vessels and Piping | Year: 2012

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487-492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. © 2011 Elsevier Ltd.


Mikus J.M.,Research Center Rez Ltd.
Nuclear Engineering and Design | Year: 2014

Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs):Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores,Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp.;Neighboring baffle in WWER-1000 type cores, andNeighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation. © 2013 Elsevier B.V.

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