Methodology of 63Cu(n,2n)62Cu Reaction Rate Measurement

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Martin Schulc ◽  
Michal Košťál ◽  
Jan Šimon ◽  
Evžen Novák
Keyword(s):  
Spontaneous Fission ◽  
Reaction Rate ◽  
Gamma Spectroscopy ◽  
Nuclear Data ◽  
Fission Neutron ◽  
Neutron Field ◽  
Gamma Line ◽  
Annihilation Peak ◽  
Standard Neutron ◽  
Data Libraries

Abstract This paper presents the measurement of the spectrum-averaged cross section (SACS) of 63Cu(n,2n)62Cu reaction in 252Cf spontaneous fission neutron spectrum. The SACS in the 252Cf spectrum was chosen as a validation tool since 252Cf is the only standard neutron field and 62Cu isotope is not easy to measure by gamma spectroscopy since the gamma line of interest is an annihilation peak, which is also produced by 64Cu isotope. Fortunately, contributions to the annihilation peak from these isotopes can be distinguished due to the very different half-lives. SACS was inferred from the experimental reaction rate. The SACS in the 252Cf spontaneous fission neutron field for the 63Cu(n,2n)62Cu reaction was determined as equal to (0.1763 ± 0.0077 mb). This value agrees with value (0.183 ± 0.007) × 10−3 b within uncertainty presented by W. Mannhart. However, it differs by 12.7% from IRDFF-II value, which is equal to (0.19874 ± 8.954 10−3) × 10−3 b. Furthermore, reasonable agreement is not achieved with ENDF/B-VIII.0, JEFF-3.3, CENDL-3.1, ROSFOND-2010, nor JENDL-4.0 nuclear data libraries.

Neutron Field Shaping Using Graphite for Reaction Rate Measurements

2021 ◽  
Vol 7 (2) ◽  
Author(s):  
Mikita Sobaleu ◽  
Michal Košťál ◽  
Jan Šimon ◽  
Evžen Losa
Keyword(s):  
Cross Section ◽  
Reaction Rate ◽  
Nuclear Data ◽  
Data File ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Neutron Field ◽  
Pure Graphite ◽  
24Na Reaction ◽  
Data Libraries

Abstract Neutron field shaping is the suitable method for validation of cross section in various energy regions. By increasing the share of neutrons of a certain energy interval and decreasing the share of other, a reaction becomes more sensitive to selected neutrons. As a result, reaction cross section can be validated in selected energy regions more precisely. The shaping can be carried out by both neutron filters which are materials with high absorption in some energy region, or by diffusion material changing the shape of neutron spectra by means of slowing down process. In the presented experiments, the neutron field of the light reactor 0 (LR-0) research reactor was shaped by both using graphite blocks inserted into the core and Cd cladding for increasing the epithermal reaction rate share in total reaction rates. The calculations were carried out with the Monte Carlo N-Particle Transport Code 6 (MCNP6) code and the most recent nuclear data libraries. The results in the pure graphite neutron field are in good agreement; in case of Cd cladding, significant discrepancies were reported. In case of the 23Na(n,γ)24Na reaction, overestimation by about 14% was reached in International Reactor Dosimetry and Fusion File (IRDFF-II), results in other libraries are comparable. In case of 58Fe(n,γ)59Fe, the overestimation as high as 18% is reported in IRDFF-II. For 64Zn(n,γ)65Zn reasonable agreement was reached in evaluated nuclear data file (ENDF/B-VIII), where discrepancies in pure graphite neutron field or in case of Cd cladding are about 10–15%.

Field Examples of Waste Assay Solutions for Curium-Contaminated Wastes

2009 ◽  
Author(s):  
Patrick M. J. Chard ◽  
Stephen Croft ◽  
Ian G. Hutchinson ◽  
Tom W. Turner ◽  
Ann Ross ◽  
...  
Keyword(s):  
Spontaneous Fission ◽  
Neutron Emission ◽  
Nuclear Data ◽  
Fission Neutron ◽  
Nuclear Material ◽  
Spent Fuel ◽  
Coincidence Counting ◽  
Plant Operations ◽  
Technical Solutions

Passive neutron coincidence counting is a mature technique for the assay of Pu in nuclear material. It is widely deployed in safeguards and waste inventory verification applications. The presence of 242Cm and 244Cm in spent fuel wastes, often poses a severe challenge owing to the relatively short spontaneous fission half-life for these isotopes and the subsequent prolific spontaneous fission neutron emission. This is a well documented problem, compounded by the fact that for most waste assay applications, neutron assay techniques are not capable of distinguishing between these Cm isotopes and the even isotopes of Pu, which are normally of interest in waste assay applications. Therefore the presence of even small quantities of these isotopes can result in gross over-estimation of the Pu inventory, if an appropriate correction is not made. Previous theoretical studies carried out recently have illustrated the potential magnitude of the problem, with reference to the fundamental nuclear data and typical isotopic compositions of wastes. Neutron multiplicity counting can, in principle, differentiate between isotopes that undergo spontaneous fission, however in practice the uncertainties in waste assay are such that this is rarely beneficial. More practical “compensation” techniques use combinations of different assay techniques (for example passive and active neutron counting) and knowledge of the actinide ratios in the waste stream fingerprint. In this paper we describe various waste assay applications as case studies. For each example we describe the nature of the challenge and show how solutions have been developed for applications where the presence of curium has caused problems. We describe the technical solutions, showing the limitations and assumptions of each. We also emphasise the role of robust Quality Assurance procedures, to ensure that the techniques are implemented reliably and with predictable outcomes. Finally, we describe the benefits that have been realised for the plant operations teams, with regard to improved measurement accuracy, avoidance of false over-estimation of the Pu inventory and subsequent improvement in plant throughput.

scholarly journals Yalina-thermal facility neutron characteristic computational study 129I, 237Np and 243Am transmutation reaction rates calculations

2020 ◽  
Vol 239 ◽  
pp. 22013
Author(s):  
Tamara Korbut ◽  
Maksim Kravchenko ◽  
Ivan Edchik ◽  
Sergey Korneev
Keyword(s):  
Monte Carlo ◽  
Computational Study ◽  
Nuclear Research ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Neutron Field ◽  
Minor Actinide ◽  
National Academy Of Sciences ◽  
Data Libraries ◽  
Academy Of Sciences

Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.

scholarly journals Gamma-ray spectroscopy measurements and simulations for uranium mining

2018 ◽  
Vol 170 ◽  
pp. 05003 ◽  
Author(s):  
T. Marchais ◽  
B. Pérot ◽  
C. Carasco ◽  
P-G Allinei ◽  
P. Chaussonnet ◽  
...  
Keyword(s):  
Gamma Ray ◽  
Uranium Concentration ◽  
Gamma Spectroscopy ◽  
Germanium Detector ◽  
Nuclear Data ◽  
Uranium Content ◽  
Atomic Data ◽  
X Rays ◽  
Data Libraries ◽  
Nuclear Data Libraries

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration evaluation by means of gamma measurements. This paper reports gamma-ray spectra, recorded with a high-purity coaxial germanium detector, on standard cement blocks with increasing uranium content, and the corresponding MCNP simulations. The detailed MCNP model of the detector and experimental setup has been validated by calculation vs. experiment comparisons. An optimization of the detector MCNP model is presented in this paper, as well as a comparison of different nuclear data libraries to explain missing or exceeding peaks in the simulation. Energy shifts observed between the fluorescence X-rays produced by MCNP and atomic data are also investigated. The qualified numerical model will be used in further studies to develop new gamma spectroscopy approaches aiming at reducing acquisition times, especially for ore samples with low uranium content.

Measurement of Selected Differential Cross Sections in 235U Spectrum

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Nicola Burianová ◽  
Michal Košt´ál ◽  
Martin Schulc ◽  
Jan Šimon ◽  
Martin Mareček ◽  
...  
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Fission Neutron ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Prompt Fission Neutron Spectrum ◽  
Prompt Fission ◽  
Prompt Fission Neutron ◽  
Good Agreement

This paper describes the measurement of 55Mn(n,2n) and 127I(n,2n) reaction rates in a well-defined reactor field in a special core of LR-0 reactor. The reaction rates were derived using gamma-spectrometry by measuring gamma activities of irradiated MnO2 and NaI samples at a high purity germanium (HPGe) detector. The spectral average cross section (SACS) in 235U prompt fission neutron spectrum (PFNS) was experimentally determined to be 0.2393 ± 0.015 × 10−3 b for 55Mn and 1.2087 ± 0.052 × 10−3 b for 127I. These obtained results were compared with calculations by MCNP6 code using ENDF/B VII.1, ENDF/B VII, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, CENDL-3.1, and IRDFF nuclear data libraries. In a case of 55Mn, a good agreement with ENDF/B VII.1, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND, and CENDL 3.1 nuclear data libraries was found, where C/E−1 is 0.1%, while IRDFF underestimated by about 15.8%. In the case of 127I, more significant discrepancies were found, where JENDL 3.3 and JENDL 4 overestimate the result by about 31.3%.

scholarly journals Comparative survey of evaluated nuclear data libraries for fusion-relevant neutron activation spectrometry

2020 ◽  
Vol 239 ◽  
pp. 21003
Author(s):  
Prasoon Raj ◽  
Ulrich Fischer ◽  
Axel Klix ◽  
JET Contributors
Keyword(s):  
Cross Section ◽  
Nuclear Physics ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Neutron Field ◽  
Physics Institute ◽  
Section Data ◽  
Fusion Applications ◽  
Data Files ◽  
Data Libraries

The neutron flux-spectrum in a fusion device is frequently determined with activation foils and adjustment of a guess-spectrum in unfolding codes. Spectral-adjustment being a rather complex and uncertain procedure, we are carefully streamlining and evaluating it for upcoming experiments. Input nuclear cross-section data holds a vital position in this. This paper presents a survey of common dosimetry reactions and available data files relevant for fusion applications. While the IRDFF v1.05 library is the recommended source, many reactions of our interest are found missing in this. We investigated other standard sources: ENDF/B-VIII.0, EAF-2010, TENDL-2017, JENDL-4.0 etc. And, we analysed two experiments to ascertain the sensitivity of the spectral adjustment to the choice of nuclear data. One was performed with D-D (approx. 2.5 MeV peak) neutrons at the Joint European Torus (JET) machine and another with a white neutron field (approx. 33 MeV endpoint energy) at Nuclear Physics Institute (NPI) of Řež. Choice of cross-section source has affected the integral fluxes (<5%), reaction rates (<10%), total fluxes in some sensitive energy-regions (>20%) and individual group fluxes (<30%). Based on this experience, essential qualitative conclusions are made to improve the fusion activation-spectrometry.

Investigation of 127I(n,2n)126I and 23Na(n,2n)22Na Reactions Using 252Cf Neutron Source

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Martin Schulc ◽  
Michal Košťál ◽  
Evžen Novák ◽  
Jan Šimon ◽  
Nicola Burianová
Keyword(s):  
Neutron Source ◽  
Cross Sections ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Fission Neutron ◽  
High Energy ◽  
High Energy Tail ◽  
252Cf Neutron Source ◽  
Data Libraries ◽  
Nuclear Data Libraries

This work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6·× 108 n/s was employed for the irradiation of sodium iodide. The spectrum-averaged cross sections (SACS) were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the net peak areas (NPAs) measured using the high purity germanium spectroscopy. The measured SACS for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to (8.98±0.32)·× 10−6 b. The resulting SACS for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as (2.044±0.0072)·× 10−3 b. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.

Criticality Calculations and Basic Sensitivity/Uncertainty Investigation of LR-0 Benchmark Core

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Tomáš Czakoj ◽  
Evžen Losa
Keyword(s):  
Lithium Fluoride ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Multiplication Factor ◽  
Reference Case ◽  
Energy Version ◽  
Data Libraries ◽  
Nuclear Data Libraries

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

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