Results of Testing the Cross Section and Related Uncertainty Data to Be Used in the New International Reactor Dosimetry File Irdf-2002

Zsolnay, E.M.; Nolthenius, H. J.

doi:10.1142/9789812705563_0063

Cited by 8 publications

(9 citation statements)

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“…The capture activity is potentially subject to contamination from lower energy, especially thermal, "room return" neutrons since the (n,γ) cross section at 25 meV is approximately 2,000 times greater than at 2.7 MeV [24,25].…”

Section: As Voyles Et Al / Nuclear Instruments and Methods In Physmentioning

confidence: 99%

“…The approach used in both measurements was to irradiate foils of zinc or titanium, which were co-loaded with indium foils in order to determine their (n,p) cross sections relative to the well-established 113 In(n,n') 113m In and 115 In(n,n') 115m In neutron dosimetry standards [24,25]. Table 1 lists physical characteristics of each foil for the various irradiations.…”

Section: Cross Section Determination By Relative Activationmentioning

confidence: 99%

“…The largest source of systematic uncertainty in the cross section determined via the "ratio approach" is the 2.586% uncertainty in the 115 In(n,n') 115m In cross section and the 1.447% uncertainty in the 113 In(n,n') 113m In cross section [24,25]. An additional uncertainty arises from the fact that the Zn/Ti samples are not located at exactly the same location as the indium monitor foils, and are therefore not subject to precisely the same neutron flux.…”

Section: Systematic Uncertaintiesmentioning

confidence: 99%

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Measurement of the 64Zn,47Ti(n,p) cross sections using a DD neutron generator for medical isotope studies

Voyles¹,

Basunia²,

Batchelder³

et al. 2017

Nuclear Instruments and Methods in Physics Research Section B:

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Cross sections for the 47 Ti(n,p) 47 Sc and 64 Zn(n,p) 64 Cu reactions have been measured for quasi-monoenergetic DD neutrons produced by the UC Berkeley High Flux Neutron Generator (HFNG). The HFNG is a compact neutron generator designed as a "flux-trap" that maximizes the probability that a neutron will interact with a sample loaded into a specific, central location. The study was motivated by interest in the production of 47 Sc and 64 Cu as emerging medical isotopes. The cross sections were measured in ratio to the 113 In(n,n') 113mIn and 115 In(n,n') 115mIn inelastic scattering reactions on co-irradiated indium samples. Post-irradiation counting using an HPGe and LEPS detectors allowed for cross section determination to within 5% uncertainty. The 64 Zn(n,p) 64Cu cross section for 2.76 +0.01−0.02 MeV neutrons is reported as 49.3 ± 2.6 mb (relative to 113 In) or 46.4 ± 1.7 mb (relative to 115 In), and the 47 Ti(n,p) 47 Sc cross section is reported as 26.26 ± 0.82 mb. The measured cross sections are found to be in good agreement with existing measured values but with lower uncertainty (<5%), and also in agreement with theoretical values. This work highlights the utility of compact, flux-trap DD-based neutron sources for nuclear data measurements and potentially the production of radionuclides for medical applications.

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Section: As Voyles Et Al / Nuclear Instruments and Methods In Physmentioning

confidence: 99%

Section: Cross Section Determination By Relative Activationmentioning

confidence: 99%

Section: Systematic Uncertaintiesmentioning

confidence: 99%

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Measurement of the 64Zn,47Ti(n,p) cross sections using a DD neutron generator for medical isotope studies

Voyles¹,

Basunia²,

Batchelder³

et al. 2017

Nuclear Instruments and Methods in Physics Research Section B:

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“…The NJpp code was developed to read NJOY99 output files and reformat them into nuclear data libraries for subsequent use by STAYSL PNNL and SHIELD (as discussed later). Future work is planned to update the nuclear data libraries with the IRDFF (International Reactor Dosimetry and Fusion File), recently released by the IAEA [4].…”

Section: Nuclear Cross Sections and Covariancesmentioning

confidence: 99%

“…STAYSL PNNL is provided as part of a comprehensive suite of programs [2], where additional tools in the suite are used for assembling a set of nuclear data libraries and determining all required corrections to the measured data to determine saturated activation rates. Neutron cross section and covariance data are taken from the International Reactor Dosimetry File (IRDF-2002) [3], which was sponsored by the International Atomic Energy Agency (IAEA), though work is planned to update to data from the IAEA's International Reactor Dosimetry and Fusion File (IRDFF) [4]. The nuclear data and associated covariances are extracted from IRDF-2002 using the third-party NJOY99 computer code [5].…”

mentioning

confidence: 99%

Least-Squares Neutron Spectral Adjustment with STAYSL PNNL

Greenwood

Johnson

2016

EPJ Web of Conferences

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Abstract. The STAYSL PNNL computer code, a descendant of the STAY'SL code [1], performs neutron spectral adjustment of a starting neutron spectrum, applying a least squares method to determine adjustments based on saturated activation rates, neutron cross sections from evaluated nuclear data libraries, and all associated covariances. STAYSL PNNL is provided as part of a comprehensive suite of programs [2], where additional tools in the suite are used for assembling a set of nuclear data libraries and determining all required corrections to the measured data to determine saturated activation rates. Neutron cross section and covariance data are taken from the International Reactor Dosimetry File (IRDF-2002) [3], which was sponsored by the International Atomic Energy Agency (IAEA), though work is planned to update to data from the IAEA's International Reactor Dosimetry and Fusion File (IRDFF) [4]. The nuclear data and associated covariances are extracted from IRDF-2002 using the third-party NJOY99 computer code [5]. The NJpp translation code converts the extracted data into a library data array format suitable for use as input to STAYSL PNNL. The software suite also includes three utilities to calculate corrections to measured activation rates. Neutron self-shielding corrections are calculated as a function of neutron energy with the SHIELD code and are applied to the group cross sections prior to spectral adjustment, thus making the corrections independent of the neutron spectrum. The SigPhi Calculator is a Microsoft Excel spreadsheet used for calculating saturated activation rates from raw gamma activities by applying corrections for gamma self-absorption, neutron burn-up, and the irradiation history. Gamma self-absorption and neutron burn-up corrections are calculated (iteratively in the case of the burn-up) within the SigPhi Calculator spreadsheet. The irradiation history corrections are calculated using the BCF computer code and are inserted into the SigPhi Calculator workbook for use in correcting the measured activities. Output from the SigPhi Calculator is automatically produced, and consists of a portion of the STAYSL PNNL input file data that is required to run the spectral adjustment calculations. Within STAYSL PNNL, the least-squares process is performed in one step, without iteration, and provides rapid results on PC platforms. STAYSL PNNL creates multiple output files with tabulated results, data suitable for plotting, and data formatted for use in subsequent radiation damage calculations using the SPECTER computer code (which is not included in the STAYSL PNNL suite). All components of the software suite have undergone extensive testing and validation prior to release and test cases are provided with the package. a Corresponding author: larry.greenwood@pnnl.gov This is an Open Access article distributed under the terms of the Creative Commons Attribution License 2.0, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

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Measurement of the 232Th(n,γ)233Th and 232Th(n,2n)231Th reaction cross-sections at neutron energies of 8.04±0.30 and 11.90±0.35MeV

Crasta

Naik

Suryanarayana

et al. 2012

Annals of Nuclear Energy

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Results of Testing the Cross Section and Related Uncertainty Data to Be Used in the New International Reactor Dosimetry File Irdf-2002

Cited by 8 publications

References 0 publications

Measurement of the 64Zn,47Ti(n,p) cross sections using a DD neutron generator for medical isotope studies

Measurement of the 64Zn,47Ti(n,p) cross sections using a DD neutron generator for medical isotope studies

Least-Squares Neutron Spectral Adjustment with STAYSL PNNL

Measurement of the 232Th(n,γ)233Th and 232Th(n,2n)231Th reaction cross-sections at neutron energies of 8.04±0.30 and 11.90±0.35MeV

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