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This paper deals with the measurement of Spectrum Averaged Cross Sections in two different neutron fields formed in zero power reactors. The first was Benchmark Neutron Reference Field in the LR-0 reactor, and the second field was in the center of the vertical channel touching the fuel in the VR-1 reactor. The spectrum averaged cross section differs for both cases as the spectra differ, but after normalization to 235U PFNS using calculated correction, both results are in good agreement, thus confirming the spectra in both cases are similar in the 1 – 14 MeV region. A good agreement between lower threshold reactions averaged in actual reactor spectra and prompt fission neutron spectrum of 235U is reported as well. [ABSTRACT FROM AUTHOR]

The assessment of the neutron flux distribution in nuclear power plant components, like reactor internals and the pressure vessel, is one of the most important parts of their residual lifetime evaluation process because the irradiation damage of these components is strongly dependent on it. The fast neutron fluences in power reactors are generally determined using calculations and verified by measurements. Discrepancies between them sometimes occur, which can be caused by be the inaccuracies in the power distribution used in neutron transport calculations. This paper quantifies the effect of uncertainty in power density on the neutron fluences behind the reactor pressure vessel (RPV). An increase in power density was experimentally simulated by implementing a higher uranium enrichment of certain selected pins. The experiment was also simulated using the MCNP code with the ENDF/B-VII.1 library. Both the experimental and calculational data show a significant local increase in neutron flux. Even behind the RPV, the increase is as high as 25%, while the local power increase in the perturbed pins is about 70–80%, which is approximately equal to the 2 σ power density uncertainties. A good agreement between measurement and calculation was found. [ABSTRACT FROM AUTHOR]

Prompt gamma radiation from neutron interactions is an important issue as it affects the operation of nuclear facilities (radiation protection or gamma heating issues) or has various uses (non-destructive identification of elements). Despite of importance, its production is inaccurately described in present nuclear data libraries. Therefore, a set of experiments have been carried out focusing on this radiation - prompt gamma from neutron inelastic scattering on oxygen and prompt gamma from radiative capture on manganese, both present the form of an aqueous solution. The gammas were induced by the Am-Be neutron source (for oxygen) and 252Cf spontaneous fission neutron source (for manganese). In the case of oxygen, the present libraries can be used, but all the libraries tested overestimated the experiment by 20–30%, considering the experimental uncertainty of 6-9%. INDEN-4.0 library is giving the best results. In the case of manganese, none of the libraries tested give acceptable results. The INDEN updated data for gamma production on manganese give usable results, but there is still a discrepancy from -36% to 55% with uncertainties from 6-15%. [ABSTRACT FROM AUTHOR]

Liquid organic scintillators are important devices for measurements of neutron radiation. This work aims to develop and optimize the composition of liquid organic scintillators so it can be used for fast neutron spectrometry. As the neutron radiation is usually accompanied with γ ray radiation, it is important to have quality γ/n discrimination. The new cocktail for house made liquid organic scintillator is prepared and studied with intention of being able to separate gamma and neutron for neutron energies above 0.5 MeV while keeping lower constraints on practical use (e.g., sealing because of oxygen) than commercial liquid scintillators. In preceding work the composition of liquid scintillators was optimized. Two twocomponent scintillators were selected for further studies. Solvent DIPN (Di-iso-propyl-naphthalene Mixed Isomers) is selected for both. First is mixed with luminophore PYR (1-Phenyl-3-(2,4,6-trimethyl-phenyl)-2-pyrazoline) of concentration 5 g/l. Second is mixed with luminophore THIO (2,5-Bis(5-tert-butyl-benzoxazol-2-yl)thiophene) of concentration 5 g/l. In this work the response of scintillator to monoenergetic beam of neutrons was measured for multiple neutron energies at PTB in Braunschweig. The two parameter spectrometric system NGA-01 is used to analyze the energy and discrimination characteristics. 137 Cs and 60 Co are used as radiation sources for calibration with pure γ rays. Then the response of scintillator for same neutron energies was simulated using GEANT4. The dissipated energy in the scintillator in response to monoenergetic neutrons is obtained. Both, measured and simulated responses are compared. Functional dependence for yield of recoiled products is estimated. It is seen that main recoil product hydrogen proton is well observed in both. From the edge of proton response one can assume the yield for given neutron energy. The recoiled carbon ion (from elastic collision) is on the other side difficult to observe in measured results but clearly seen in dissipated energy plots. It suggests that yield of carbon ion is very small relatively to proton yield. These results will serve as basis for response function evaluation of scintillator which is necessary for evaluation of unknown neutron spectra from measurements with scintillator. [ABSTRACT FROM AUTHOR]

In this paper we study quality of three types of organic scintillators stilbene, p-terphenyl and EJ-299-33. We used monoenergetic neutron fields with a wide range of neutron energies in PTB Braunschweig. All the tests were carried out with NGA-01 spectrometer. The results of the measurements are evaluated spectra from the spectrometer. We discuss the quality of each scintillator, such as the FWHM of the peaks. [ABSTRACT FROM AUTHOR]

Pool-type reactors are an excellent source of high energy photons without neutrons because the water above fuel acts as an outstanding neutron shielding, whereas for high energy photons, it is nearly transparent. The absence of neutrons is a valuable property essential for testing semiconductor detectors because they are sensitive to neutrons, which might cause their damage. The high energy gammas are worth studying because they are responsible for radiation induced heating of reactor internal components in energy producing reactors, which might lead to their radiation induced damage by void swelling. At this effect, the helium produced in (n,α) reactions influenced by locally increased temperature clusters into bigger bubbles, which are increasing material volume, and therefore changing the geometry of internals. The referred method could be utilized e.g. for resource exploration by detecting prompt capture gammas. The prompt gamma field was measured at the VR-1 reactor with two different cores and water levels. The stilbene measurements showed relatively good agreement with the spectrum shape computed by MCNP6.2. An important result is a confirmation that the stilbene deconvolution matrix can be used with a rotated crystal. The testing measurement with HPGe detectors showed a possibility for measurement in such type of geometry. [ABSTRACT FROM AUTHOR]

Silicon cross-sections, being important for a criticality safety of final spent fuel disposals, were recently reevaluated within the IAEA INDEN project. Similarly, the thermal Scattering Law matrix for silicon dioxide, which is also important for criticality safety, was also reevaluated in the ENDF/B-VIII.0 library. Due to these reasons, a series of validation experiments with silica sand were performed at the LR-0 reactor. This paper describes these validation experiments, which used two different amounts of silica sand placed in the core. The first part of validation was carried out as critical experiments in order to benefit from the suitability of integral experiments for validation. The second part -fast neutron spectrum measurement in the sand -was performed in order to obtain knowledge of its characteristics and its agreement with a calculation. The results showed significant improvement of the Thermal Scattering Law matrix for silicon dioxide available in the ENDF/B-VIII.0 library. They also showed that the new INDEN evaluation of silicon cross-sections, together with its description in the ENDF/B-VIII.0 gives disagreement rate closest to the experiments carried out without silica sand insertions. The spectrum measurement showed that the calculations of fast neutron spectra in the sand show only slight differences between different evaluations of silicon cross-sections. However, the fast neutron spectrum is not dependent on the Thermal Scattering Law. The calculated spectra show relatively good agreement with the measurement. [ABSTRACT FROM AUTHOR]

Experimental work at the LR-0 reactor was carried out to determine the possibility of the mock-up neutron field creation for the fluoride salt-based reactors. Previous calculations and experiments have shown that the fast part of the molten salt reactor (MSR) spectrum is shaped by fluorine and even the Teflon material is suitable for neutronics in that energy range. Properly selected spectrum indices can describe the neutron field of the MSR in fast thermal and intermediate parts of the spectrum. Current research has focused on a deeper study of the possibility of using the filtered thermal and intermediate neutron spectrum of the experimental light water reactor for the physics of fluoride salt-based reactor. LiF-BeF2 (FLIBE) capsule and teflon cylinders are used as a spectrum filters in the LR-0 reactor. Measured results show acceptable C/E-1 agreement in the reaction rates and satisfactory agreement for usage of the FLIBE filtered neutron spectrum determined by the 181Ta(n, γ) monitor as a mock-up in thermal to intermediate energy in the fluoride high-temperature reactor (FHR). Concerning MSR, intermediate spectrum can be reproduced to some extent but not as good as in the case of FHR. [ABSTRACT FROM AUTHOR]

In this work we compare the pulse-shape discrimination (PSD) properties of EJ-299-33A, BC-501A, stilbene, p-terphenyl and Hidex Aqualight in neutron field generated by the LVR-15 reactor with silicon filter utilization. Pulses from the scintillators are processed by Neutron-Gamma spectrometer. This spectrometric system with fast digitizer card contains two analog-digital (A/D) converters with a resolution of 12 bits and sampling frequency 500 MHz. For photomultiplier (PMT) linearity improvement active divider has been used. Measured data from scintillators have been processed using the integration method and compared. Results are presented. [ABSTRACT FROM AUTHOR]

The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor. [ABSTRACT FROM AUTHOR]

The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures. [ABSTRACT FROM AUTHOR]

The presented paper aims to compare the calculated and experimental reaction rates of 23Na(n,2n)22Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination. Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of longterm activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. In the case of reactor spectrum, reasonable agreement was not achieved with any library. However, in the case of 252Cf spectrum agreement was achieved with IRDFF, JEFF-3.1 and JENDL libraries. [ABSTRACT FROM AUTHOR]