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![Few examples of calculated uncertainties (in pcm) in [34] from covariance files.](publication/319672764/figure/tbl2/AS:631646629924919@1527607700926/Few-examples-of-calculated-uncertainties-in-pcm-in-34-from-covariance-files.png)
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![Table 2 . Few examples of calculated uncertainties (in pcm) in [34]...](publication/319672764/figure/tbl2/AS:631646629924919@1527607700926/Few-examples-of-calculated-uncertainties-in-pcm-in-34-from-covariance-files_Q320.jpg)
The challenge for any nuclear data evaluation project is to periodically release a revised, fully consistent and complete library, with all needed data and covariances, and ensure that it is robust and reliable for a variety of applications. Within an evaluation effort, benchmarking activities play an important role in validating proposed libraries...
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... and JENDL-4.0 are tested in the NEA criticality validation suite. As it was expected, large differences in keff uncertainties were found due to 239 Pu and 235 U covariance data in JEFF-3.3.T1 (see Table 2). Sensitivity and uncertainty analysis for the Tritium Breeding Ratio (TBR) of a DEMO Fusion Reactor was carried out with different sets of covariance data: JEFF-3.2, ...
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Citations
... Other evaluations used in this paper, i.e., the JEFF-3.3 [31] or the ENDF/B-VII.1 [32] evaluations, were based on the Los Alamos or Madland-Nix model [33]. ...
... The transport files employed in the transport and correction calculations were taken from the ENDF/B-VII.I library [32] except for 235 U which was taken alternatively from ENDF/B-VIII.0 library [26], ENDF/B-VII.1 library, or JEFF-3.3 [31] to show major differences in evaluated prompt fission thermal-neutron spectrum. ENDF/B-VII.I library was chosen for transport because of consistency with older calculations before the release of the ENDF/B-VIII.0 ...
Spectrum-averaged cross-section (SACS) ratio measurement showing reduced uncertainties of measured SACS in two independent fission neutron fields is presented. The used prompt fission neutron fields correspond to the 252Cf(sf) and 235U(nth,f) fission neutrons. The employed SACS in the 235U(nth,f) prompt fission neutron spectrum (PFNS) were measured using three different light water reactors: LR-0 and VR-1 zero power reactors and the LVR-15 10 MWt reactor. The employed SACS in the 252Cf(sf) prompt fission neutron field were measured using a certified high-flux neutron source. Existing correlations among measured SACS in the two different neutron fields are estimated and used to reduce the uncertainty of measured SACS ratio for IRDFF dosimetry reactions. The derived set of measured SACS ratios with reduced uncertainty extends previous works into the higher-energy fission neutron range up to 20–30 MeV. The SACS ratio in 252Cf(sf) and 235U(nth,f) PFNS can be used to probe the high-energy tail of the 235U(nth,f) fission neutron spectrum as the 252Cf(sf) reference neutron spectrum is relatively well known. Derived experimental SACS ratio data. featuring low uncertainty are compared to the calculated dosimetry SACS ratio using the IRDFF-II dosimetry cross sections and the 235U(nth,f) ENDF/B-VII.1, ENDF/B-VIII.0, or JEFF-3.3 PFNS evaluations. The ENDF/B-VIII.0 evaluation of the 235U(nth,f) PFNS agrees well with derived SACS ratio data within quoted uncertainties. Other libraries predict a significantly lower fraction of 235U(nth,f) fission neutrons above 11 MeV of the outgoing neutron energy.
... and JEFF-3.2 [[5], [6], [7], [8], [9]], and to compare the results with the references. The OECD/NEA Expert Group proposed the VVER-1000 LEU and MOX assembly computational benchmark as a theoretical benchmark for investigating the physics of a whole VVER-1000 reactor core using two-thirds (LEU) and one-third (MOX) fuel [ [10], [11]]. ...
This article analyses the OpenMC Monte Carlo code's ability to predict eigenvalue values for low-enriched uranium (LEU) and weapons-grade (MOX) fuel assemblies of the Russian VVER-1000 reactor. Besides, the differences in performance of the five most recent data libraries, i.e. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 have been evaluated by comparing the resulting reactivities of three calculation states, namely soluble boron effect, Fuel temperature (Doppler) effect, and total temperature effect.
... The black (solid), orange (dashed), pink (dotted), blue (dash-dotted), and green (densely dashed) lines, respectively, denote CENDL-3.2 [19], ENDF/B-VIII.0, JEFF3.3 [20], BROND-3.1 [21], and JENDL-4.0HE evaluations. ...
The neutron-induced fission cross section of Pu239 was measured relative to U235(n,f) at the back-streaming white neutron beam line (Back-n) of the China Spallation Neutron Source (CSNS). A multicell fast fission ionization chamber was used to perform the measurement. The reliability of the measurement was verified by the high consistency of U235's resonances between the measurement and the evaluation data. The Pu239 fission cross sections from 4keV to 100MeV are obtained with 1.7–5.8 % uncertainty when unfolding uncertainties are excluded. The total uncertainties including the unfolding errors, which reflect the effect of the double-bunch unfolding method, are 2.6–15 % from 10keV to 100MeV.
... Figure 5(a) shows the presented FPY results and total uncertainties along with the evaluated nuclear data and experimental data in Table 2. Figure 5(b) shows that the majority of discrepancies between evaluations ENDF/B-VIII.0 [29] and JEFF-3.3 [30] are 0.1% -40%. ...
The fission yield data in 14 MeV energy neutron induced fission of 238U play an important role for decay heat calculation and for generation-IV reactor designs. In order to measure fission product yields of 238U induced by 14 MeV neutrons accurately, the cumulative yields of fission products ranging from 92Sr to 147Nd in 238U(n, f) with 14.7 MeV neutron have been determined using an off-line γ-ray spectrometric technique. The 14.7 MeV quasi-monoenergetic neutron beam was provided by K-400 D-T neutron generator at China Academy of Engineering Physics (CAEP). Fission products were measured by a low background high purity germanium gamma spectrometer. The neutron flux was obtained from 93Nb (n,2n)92mNb reaction, and the mean neutron energy was calculated by the cross-section ratios for the 90Zr(n,2n)89Zr and 93Nb(n,2n)92mNb reactions. With a series of corrections, high precision cumulative yield of 19 fission products were obtained. Our FPYs for 238U(n,f) reaction at 14.7 MeV were compared with the existing data at 14 MeV from Experimental Nuclear Reaction Data and Evaluated Nuclear Data, respectively. They are helpful to design of generation-IV reactor and the construction of evaluated fission yield database.
... However, ENDF/B-VIII.0[19] and JEFF-3.3[20] have their own dedicated decay libraries. ...
A reliable estimation of the operational parameters is one of the primary concerns in the design of magnetic fusion devices such as ITER and DEMO. Methods of diagnostics and control over the critical plasma parameters determining its stability and efficiency rely on the high-energy neutron field monitoring. Extreme operational environment, such as high-energy neutron flux, electromagnetic radiation, and high temperatures might reduce the performance of the detector systems. Therefore, research and development activities in detector prototyping are carried out to address this problem. To predict the performance of the detector materials, simulations using the latest releases of the nuclear data libraries as input for the inventory codes are carried out. This paper describes the latest validation and verification (V&V) benchmark exercise for FISPACT-II & TENDL-2017 based on the fusion decay heat measurements performed at the Japanese FNS facility for the materials in the diagnostic components for the radiation measurements. The breakdown of decay-heat contributions from individual radionuclides have been employed to interpret the simulated results, benchmark the data against the experimental measurements, and revise the neutron-induced reactions cross-section and decay data for the associated radionuclides for the upcoming release of the TENDL-2019 nuclear data library.
... In the Evaluated Nuclear Data File (ENDF; Trkov and Brown, 2018), thermal neutron scattering in an atomic system is explained using the thermal scattering law (TSL), which is the fundamental data to evaluate thermal scattering cross-sections. Currently, individual TSL data files are contained in the major ENDF such as ENDF/B , JEFF (Cabellos et al., 2017), and JENDL (Shibata et al., 2011). However, a limited number of materials have TSL data. ...
... library contains only 33 evaluations for 24 moderator materials, whereas the JEFF-3.3 (Cabellos et al., 2017) contains only 20 evaluations for 16 materials; this does not satisfy various demands for the application of nuclear technologies. Moreover, some new moderator materials applied in advanced reactors do not have TSL data such as ZrC and UC in ultra-high temperature reactors (Difilippo and Renier, 2007), LiH in nuclear propulsion particle bed reactors (Wang et al., 2013), FLiBe in molten salt reactors (Mei et al., 2013;Zhu and Hawari, 2017), and heavy paraffinic molecular materials (Manring and Hawari, 2019). ...
A new module designed to generate thermal scattering law (TSL) data is developed in the nuclear data processing code NECP-Atlas, which is currently under development at Xi’an Jiaotong University. The module can calculate TSL data for solid materials of any crystal structure and liquid materials in the free gas or the Egelstaff and Schofield approximations. To acquire a precise description of the neutron thermalization, the module eliminates the cubic approximation and atom site approximation used in the conventional method. In addition, a technique based on the anisotropic displacement parameters method is implemented to consider the anisotropy of interatomic forces and the correlation of forces from different directions. Comparison with the LEAPR and the Debye–Waller matrix method used in ENDF/B-VIII.0 enabled verification of the TSL data generated by the module. The numerical results revealed the good performance of the module.
... However, most multigroup cross section libraries are based on the versions of ENDF/ B-VII.0 (Chadwick et al., 2006), JEFF-3.3 (Cabellos et al., 2017), JENDL-4.0 (Shibata et al., 2011), BROND-3.1 (Blokhin et al., 2016), and CENDL-3.1 (Ge et al., 2011) evaluated nuclear data libraries. ...
The aim of this paper is to present the critical verification for TRX and BAPL benchmarks between the ENDF/B-VIII.0 and ENDF/B-VII.0 library. Using DRAGON5.0.6 code with 281-group WIMS-D and DRAG library, the integral parameters for TRX and BAPL benchmarks are compared to the Monte-Carlo code and experimental values. For the verification of DRAGON code, the influence of tracking method, resonance self-shielding method, and leakage model are investigated in this paper. The results indicated that the combination methods of collision probability method, equivalence theory, and P1 leakage model have better accuracy in TRX and BAPL benchmarks. For the comparison of ENDF/B-VII.0 and ENDF/B-VIII.0 library, the difference of integral parameters is small between the ENDF/B-VII.0 and ENDF/B-VIII.0 library. The relative error of ρ²⁸ and δ²⁵ is less than 1%. However, the accuracy of δ²⁵ and C* is underestimated by about 1.0%∼1.5% in ENDF/B-VIII.0 library. It means that the fission cross section for U-235 is overestimated in ENDF/B-VIII.0 library and the 2-group cross section comparison is consistent with this tendency. Therefore, the absorption cross section for whole region is increased in thermal energy group, which finally leads to the keff and kinf decreased in ENDF/B-VIII.0 library. Compared with the experimental values in TRX and BAPL cases, the integral parameters in ENDF/B-VIII.0 are more accurate than the ENDF/B-VII.0 library.
... Visual representation of detectors inside the JSI TRIGA model. data libraries, JEFF 3.3 (Cabellos et al., 2017) nuclear data library was used for isotopic inventory and delayed radiation spectra calculations. Delayed electron source terms were calculated by the BetaShape code, with a 10 keV resolution. ...
Interest on development of computational tools for calculations of radiation field due to decay of radioactive activation and fission products (delayed radiation) has increased in recent years, mainly due to requirements on shutdown dose rate for the future ITER fusion reactor and material testing reactors (MTR), where gamma heating values of several 10 Wg−1 can be reached in reactor structural components during reactor steady state operation and for characterization of research reactor irradiation facilities. Previous experimental work shows a roughly 30% contribution of delayed gamma rays to the total gamma ray flux.
An extensive experimental campaign at the JSI TRIGA reactor has been carried out using multiple ionization and fission chamber detectors for gamma and neutron field characterization. A series of reactor power steps was performed with said detectors in different irradiation positions during steady reactor power and after shutdown. The reactor and the used detectors, along with irradiation power steps have been modelled in detail using the JSIR2S code system framework to compare measured and calculated fluxes and dose rates. Simulated results are in good agreement with measurements and with experimentally obtained values of delayed gamma fractions from previous work.
... 2018), JEFF-3.3(Cabellos et al., 2017), JENDL-4.0(Shibata et al., 2011),CENDL-3.1Ge et al., 2011, ROSFOND-2010(Zabrodskaya et al., 2007,EAF-2010 (Packer andSublet, 2010), TENDL-2017Koning andRochman, 2012 and IRDFF-II Trkov et al., 2020nuclear data libraries for comparison. The input 252 Cf(sf) neutron spectrum was taken from the IAEA IRDFF webpage. There is a possibility to ...
Only neutron spectrum standard is ²⁵²Cf spontaneous fission neutron spectrum. However, the high energy tail of this spectrum is loaded with high uncertainty. To reduce this uncertainty, it is crucial to use validated cross sections with low uncertainty. The explored set of reactions covers ⁵⁸Ni(n,X)⁵⁷Co, ¹⁶⁹Tm(n,3n)¹⁶⁷Tm, ¹⁹⁷Au(n,3n)¹⁹⁵Au, ²⁰⁹Bi(n,3n)²⁰⁷Bi and ²⁰⁹Bi(n,4n)²⁰⁶Bi threshold reactions. Measurement of dosimetric ¹⁶⁹Tm(n,3n)¹⁶⁷Tm, ²⁰⁹Bi(n,3n)²⁰⁷Bi and ²⁰⁹Bi(n,4n)²⁰⁶Bi reactions spectral averaged cross sections (SACS) is included in NEA's High Priority Nuclear Data Request List. All these reactions SACS were measured for the first time. All SACS were derived from experimentally determined reaction rates by gamma spectrometry using the same high-purity germanium detector. In the case of ¹⁶⁹Tm(n,3n)¹⁶⁷Tm reaction, the difference between experimental and calculated value using the IRDFF-II library is only 1.45%. Concerning ¹⁹⁷Au(n,3n)¹⁹⁵Au reaction, the reasonable agreement is achieved only using the TENDL-2017 library. In the case of ²⁰⁹Bi(n,3n)²⁰⁷Bi reaction, agreement within uncertainty is not achieved with any library unlike ²⁰⁹Bi(n,4n)²⁰⁶Bi reaction where the agreement within uncertainty is achieved with IRDFF-II library. The best agreement for ⁵⁸Ni(n,X)⁵⁷Co reaction is achieved using ENDF/B-VIII library.
... without further change. A summary of benchmarking and validation activities for the first test releases of JEFF-3.3 can be found in Reference [368]. ...
... As an outcome of the dedicated JEFF group for processing and verification [368], the following standard sequence of NJOY modules was set up and optimal input parameters were identified to process the JEFF-3.3 incident-neutron data into ACE format: RECONR: the ENDF-6 file is converted into pointwise-ENDF (PENDF) format, where sufficient energy grid points are introduced to ensure that linear-linear interpolation reproduces the original evaluations within an accuracy of 0.1%. ...
The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides \(^{235}\hbox {U}\), \(^{238}\hbox {U}\) and \(^{239}\hbox {Pu}\), on \(^{241}\hbox {Am}\) and \(^{23}\hbox {Na}\), \(^{59}\hbox {Ni}\), Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.
