Filters
Results 1 - 10 of 33
Results 1 - 10 of 33.
Search took: 0.035 seconds
| Sort by: date | relevance |
[en] In this work, an overview on the relevance of the nuclear data (ND) uncertainties with respect to the Light Water Reactors (LWR) neutron dosimetry is presented. The paper summarizes results of several studies realized at the LRT laboratory of the Paul Scherrer Institute over the past decade. The studies were done using the base LRT calculation methodology for dosimetry assessments, which involves the neutron source distribution representation, obtained based on validated CASMO/SIMULATE core follow calculation models, and the subsequent neutron transport simulations with the MCNP® software. The methodology was validated using as reference data results of numerous measurement programs fulfilled at Swiss NPPs. Namely, the following experimental programs are considered in the given overview: PWR “gradient probes” and BWR fast neutron fluence (FNF) monitors post irradiation examination (PIE). For the both cases, assessments of the nuclear data related uncertainties were performed. When appropriate, a cross-verification of the deterministic and stochastic based uncertainty propagation techniques is provided. Furthermore, the observations on which particular neutron induced reactions contribute dominantly to the overall ND-related uncertainties are demonstrated. The presented results should help with assessing the overall impact of the various nuclear data uncertainties with respect to dosimetry applications and provide relevant feedback to the nuclear data evaluators.
Journal
EPJ. Web of Conferences; ISSN 2100-014X; 
; v. 239; vp
; v. 239; vp
[en] Thermal scattering data for hydrogen in light water (H in H2O) are one of the major components of nuclear data affecting reactor and fuel calculations. The impact of the H in H2O thermal scattering data are calculated for burnup quantities, considering models of a UO2 pin cell with DRAGON (a deterministic code) and SERPENT (a Monte Carlo code). The Total Monte Carlo method is applied, where the CAB model parameters are randomly varied to produce sampled (random) LEAPR input files for NJOY. A large number of burnup calculations is then performed, based on the random thermal scattering data. It is found that the impact on k∞ is relatively small (less than 35 pcm), as for nuclide inventory (less than 1% at 50 MWd/kgU) and for decay heat (less than 0.4%). It is also observed that the calculated probability density functions indicate strong non-linear effects.
Journal
EPJ Nuclear Sciences and Technologies; ISSN 2491-9292; ; v. 7; p. 24.1-24.8
; v. 7; p. 24.1-24.8
[en] For irradiations in a materials test reactor, the prediction of the amount of gamma heating in the reactor is important. Only a good predictive calculation will lead to an irradiation in which the specified temperatures are reached. The photons produced by fission product decay are often missing in spectrum calculations for a reactor, but the contribution of the photons can be computed effectively using engineering correlations for the amount of fission product decay and the ensuing photon spectrum. The prompt photons are usually calculated by a spectrum code based on the underlying nuclear data libraries. For most of the important nuclides, the nuclear data libraries contain data for the photon productions rates. However, there are still many nuclides for which the photon production data are missing, and some of these nuclides contribute to gamma heating. In this paper it is estimated what the contributions to heating are from photon production on nuclides such as 236U, 238Pu, 135I, 135Xe, 147Pm, 148Pm, 148mPm, and 149Sm. Also, simple arguments are given to judge the effect from photon production on all other (lumped) fission products, and from 28Al decay. For all these calculations the High Flux Reactor is used as an example. (authors)
Source
2008; 5 p; Paul Scherrer Institut - PSI; Villigen PSI (Switzerland); PHYSOR'08: International Conference on the Physics of Reactors 'Nuclear Power: A Sustainable Resource'; Interlaken (Switzerland); 14-19 Sep 2008; ISBN 978-3-9521409-5-6; ; Country of input: France; 3 refs.; proceedings are available as a CD-ROM on request to info'at'physor08.ch
; Country of input: France; 3 refs.; proceedings are available as a CD-ROM on request to info'at'physor08.ch
[en] The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the samples are also provided. For uncertainty propagation, the covariance information from 3 different nuclear data libraries are used. Uncertainties from manufacturing tolerances and operating conditions are also considered. The results from these two samples are compared with the ones from two UO2 samples, namely GU1 and GU3, also from the ARIANE program, applying the same calculation scheme and uncertainty assumptions. It is shown that a two-dimensional assembly model provides better agreement with the measurements than a two-dimensional single pin model, and that the full core three-dimensional model provides similar results compared to the assembly model, although no 148Nd normalization is applied for the full core model. For the MOX assembly decay heat, as expected, heavy actinides have a higher contribution compared to the cases with the UO2 samples; additionally, decay heat uncertainties are moderately smaller in the case of the MOX assembly. (authors)
Journal
EPJ Nuclear Sciences and Technologies; ISSN 2491-9292; ; v. 7; p. 18.1-18.24
; v. 7; p. 18.1-18.24
[en] Multi-purpose methodical nuclear data libraries generated by TALYS/T6 as Evaluated Nuclear Data Library TENDL have been released regularly since 2008. Considerable experience has been acquired during those early production of such unique, truly multi-faceted, multi-particle nuclear data libraries based on the feedback from the developers, evaluators, processing experts and, most importantly, users. The backbone of this achievement is simple and robust: completeness, quality, upgradability and, most of all, reproducibility and methodology. Since TENDL has been seamlessly adopted by many different applications (accelerator, astrophysics, fusion, fission, medical) that require multi-faceted nuclear reaction data forms for shielding, radio-protection, transmutation, materials or earth sciences, it is felt necessary to understand its strengths and remaining weaknesses. The essential knowledge is not the TENDL library itself, but rather the necessary methods, processes, codes, tools and know-how that go into the making of every evaluations of such a library. Current efforts are focused on a proper evaluation of the underlying physics and incorporation of information and metrics into the scientific TALYS/T6 system. (authors)
Source
Aug 2018; 7 p; American Nuclear Society - ANS; La Grange Park, IL (United States); RPSD 2018: 20. Topical Meeting of the Radiation Protection and Shielding Division of ANS; Santa Fe, NM (United States); 26-31 Aug 2018; ISBN 978-0-89448-746-0; ; Country of input: France; 21 refs.; available on CD Rom from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)
; Country of input: France; 21 refs.; available on CD Rom from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)
| 1 | 2 | 3 | Next |