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.

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 Testing Oxygen Nuclear Data Libraries in Research Center Rez

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Martin Schulc ◽  
Michal Košťál ◽  
Evžen Novák ◽  
Jan Šimon ◽  
Luiz Leal
Keyword(s):  
Fast Neutron ◽  
Neutron Source ◽  
Nuclear Data ◽  
Light Water ◽  
Neutron Spectra ◽  
Active Core ◽  
252Cf Neutron Source ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Fast Neutron Spectrum

Abstract Oxygen belongs to the group of the most important isotopes in the nuclear data field. The aim of this paper is validate various oxygen nuclear data libraries in different scenarios with high content of oxygen. For this purpose, fast neutron spectra were measured by a stilbene scintillation detector in the region of 1–10 MeV in the three model cases involving 252Cf neutron source and light water reactor. The cases include measurements of leakage spectra using 252Cf neutron source placed in the centers of the light water and heavy water spheres of 0.30 m diameter. Following measurements were carried out inside the concrete biological shielding of the VVER-1000 mock-up simulator in the LR-0 reactor and in the dry channel located in the center of the special core placed in the LR-0 reactor. In the case of the special core, symmetric active core consisted of six standard fuel assemblies which surround the experimental dry module, where the fast neutron spectrum was measured. The measured neutron spectra were compared with MCNP6 transport code calculations in ENDF/B-VII.1, ENDF/B-VIII.0, JENDL-4.0, and IRSN 16O nuclear data evaluations. Experimental results for all cases follow similar trend. All considered libraries underestimate experimental measurement in the region of 3–4 MeV in all cases.

scholarly journals IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE

2021 ◽  
Vol 247 ◽  
pp. 09026
Author(s):  
A.G. Nelson ◽  
K.M. Ramey ◽  
F. Heidet
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Power Distribution ◽  
Evaluation Process ◽  
Nuclear Data ◽  
Reactor Performance ◽  
Data Set ◽  
Experimental Database ◽  
Data Libraries ◽  
Nuclear Data Libraries

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs. This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.

scholarly journals Evaluating nuclear data and their uncertainties

2018 ◽  
Vol 4 ◽  
pp. 29
Author(s):  
Patrick Talou
Keyword(s):  
Experimental Data ◽  
Uncertainty Quantification ◽  
Cross Sections ◽  
Coming Out ◽  
Nuclear Data ◽  
Covariance Matrices ◽  
Alternative Approaches ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Evaluated Nuclear Data

In the last decade or so, estimating uncertainties associated with nuclear data has become an almost mandatory step in any new nuclear data evaluation. The mathematics needed to infer such estimates look deceptively simple, masking the hidden complexities due to imprecise and contradictory experimental data and natural limitations of simplified physics models. Through examples of evaluated covariance matrices for the soon-to-be-released U.S. ENDF/B-VIII.0 library, e.g., cross sections, spectrum, multiplicity, this paper discusses some uncertainty quantification methodologies in use today, their strengths, their pitfalls, and alternative approaches that have proved to be highly successful in other fields. The important issue of how to interpret and use the covariance matrices coming out of the evaluated nuclear data libraries is discussed.

scholarly journals Infological models of the ENDF-format nuclear data

2019 ◽  
Vol 5 (1) ◽  
pp. 53-59
Author(s):  
Anatoliy G. Yuferov
Keyword(s):  
Cross Sections ◽  
Modern Science ◽  
Physical Nature ◽  
Nuclear Data ◽  
Secondary Neutron ◽  
Source Data ◽  
Presentation Formats ◽  
Hierarchical Nature ◽  
Data Libraries ◽  
Nuclear Data Libraries

Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website.

scholarly journals Validation of JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries in ANSWERS codes

2020 ◽  
Vol 239 ◽  
pp. 19001
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Glynn Hosking ◽  
Ray Perry ◽  
Simon Richards
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Reactor Physics ◽  
Group Data ◽  
Nuclear Data Library ◽  
Data Libraries ◽  
Neutron Cross Sections ◽  
Nuclear Data Libraries ◽  
Data Library

The JEFF-3.3 and ENDF/B-VIII.0 evaluated nuclear data libraries were released in December 2017 and February 2018 respectively. Both evaluations represent a comprehensive update to their predecessor evaluations. The ANSWERS Software Service produces the MONK® and MCBEND Monte Carlo codes, and the WIMS deterministic code for nuclear criticality, shielding and reactor physics applications. MONK and MCBEND can utilise continuous energy nuclear data provided by the BINGO nuclear data library and MONK and WIMS can utilise broad energy group data (172 group XMAS scheme) via the WIMS nuclear data library. To produce the BINGO library, the BINGO Pre-Processor code is used to process ENDF-6 format evaluations. This utilises the RECONR-BROADR-PURR sequence of NJOY2016 to reconstruct and Doppler broaden the free gas neutron cross sections together with bespoke routines to generate cumulative distributions for the S(α,β) tabulations and equi-probable bins or probability functions for the secondary angle and energy data. To produce the WIMS library, NJOY2016 is again used to reconstruct and Doppler broaden the cross sections. The THERMR module is used to process the thermal scattering data. Preparation of data for system-dependent resonance shielding of some nuclides is performed. GROUPR is then used to produce the group averaged data before all the data are transformed into the specific WIMS library format. The MONK validation includes analyses based on around 800 configurations for a range of fuel and moderator types. The WIMS validation includes analyses of zero-energy critical and sub-critical, commissioning, operational and post-irradiation experiments for a range of fuel and moderator types. This paper presents and discusses the results of MONK and WIMS validation benchmark calculations using the JEFF-3.3 and ENDF/B-VIII.0 based BINGO and WIMS nuclear data libraries.

Measurement and covariance analysis of 59Co(n, 2n)58Co reaction cross sections at the effective neutron energies of 11.98 and 15.75 MeV

2018 ◽  
Vol 106 (11) ◽  
pp. 877-884 ◽  
Author(s):  
Santhi Sheela Yerraguntla ◽  
Haladhara Naik ◽  
Manjunatha Karantha ◽  
Srinivasan Ganesan ◽  
Suryanarayana Venkata Saraswatula ◽  
...  
Keyword(s):  
Cross Sections ◽  
Present Experiment ◽  
Nuclear Data ◽  
Reaction Cross ◽  
Correlation Technique ◽  
Effective Neutron ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Evaluated Nuclear Data ◽  
Reaction Cross Sections

Abstract The 59Co(n, 2n)58Co reaction cross sections relative to the cross sections of the 115In(n, n′)115mIn reaction have been measured at the effective neutron energies of 11.98 and 15.75 MeV by using activation and off-line γ-ray spectrometric technique. Neutron beam used in the present experiment was generated from the 7Li(p, n)7Be reaction with the proton energies of 14 and 18 MeV at the 14UD BARC-TIFR Pelletron facility, Mumbai. We also present the covariance information by taking into account the sources of error and the correlations between the attributes influencing the measurements. The 59Co(n, 2n)58Co reaction cross sections from the present work are then compared with the values from different evaluated nuclear data libraries. The micro-correlation technique suggested by Smith was modified to generate the covariance matrix for the measurements of reaction cross sections as the efficiencies of detector for the sample and monitor are correlated.

scholarly journals Validation of heavy water cross section using AmBe neutron source

2020 ◽  
Vol 239 ◽  
pp. 18008
Author(s):  
Michal Kostal ◽  
Martin Schulc ◽  
Evzen Novak ◽  
Tomas Czakoj ◽  
Zdenek Matej ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Source ◽  
Cross Sections ◽  
Average Energy ◽  
Nuclear Data ◽  
High Energy ◽  
Suitable Candidate ◽  
Neutron Spectra ◽  
Misleading Interpretation ◽  
Neutron Leakage

Physical quantities derived from integral experiments can usually be measured much more accurately than that from differential nuclear data. The accurate knowledge of integral parameters provide excellent grounds for testing and tuning differential data such as, for instance, cross sections. Measurement of neutron leakage spectra with 252Cf neutron source located at sphere center is often used for integral experiments. While this type of experiments provide information for cross section tuning, however, care must be taken to avoid misleading interpretation, namely, at high energies due to the very low portion of high energy neutrons in 252Cf spectrum. This issue can be alleviated by the use of point source with different spectra shape. For that purpose one suitable candidate seems to be the AmBe neutron source which has a relatively high average energy and peak character of emitted neutrons. Indeed, AmBe seems an interesting option because the calculated leakage neutron spectra are not very sensitive to the input shape of the neutron spectra. Thus the neutron leakage spectra calculated using tabulated of International Organization for Standardization spectra is nearly the same as stilbene measured AmBe spectra as an input.

Sign in / Sign up

Export Citation Format

Share Document