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No abstract available
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AO «Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii — Fiziko-Ehnergeticheskij Inst. imeni A.I. Lejpunskogo», Obninsk (Russian Federation); 78 p; ISBN 978-5-907108-06-6; 
; 2018; p. 58-59; Scientific-technical conference 'Neutron-physical problems of nuclear power'; Nauchno-tekhnicheskaya konferentsiya «Nejtronno-fizicheskie problemy atomnoj ehnergetiki»; Obninsk (Russian Federation); 28-30 Nov 2018; 4 refs.
; 2018; p. 58-59; Scientific-technical conference 'Neutron-physical problems of nuclear power'; Nauchno-tekhnicheskaya konferentsiya «Nejtronno-fizicheskie problemy atomnoj ehnergetiki»; Obninsk (Russian Federation); 28-30 Nov 2018; 4 refs.
No abstract available
Journal
[en] Highlights: • Analysis of SPERT-III Experiments. • Reactivity Initiated Accident Analysis. • Nuclear Data Uncertainty Propagation in CASMO down to 3-D core transient simulation. • Transient Analysis of RIA using SIMULATE-3K. • Uncertainty in Dynamical parameters. - Abstract: This research aims at validating the SIMULATE-3K code and complementing the results with the quantification of nuclear data uncertainties against the Special Power Excursion Reactor Test III (SPERT-III) experiments. To that aim, the SHARK-X methodology, under development at PSI, for the propagation of nuclear data uncertainties in CASMO5 2-D lattice calculations to 3-D core transient simulations is applied for the analysis of a SPERT-III super-prompt critical test conducted at cold startup conditions. Concerning transient results, both total power and reactivity show a good agreement with the measurements at the initial phase of power excursion, while a slight discrepancy is obtained at the final phase of the transient. The estimated uncertainties regarding both steady-state parameters such as k-eff and static reactivity worth, as well as dynamical quantities such as power pulse width and enthalpies are presented. Results show non-negligible sensitivity upon the employed nuclear data library. The uncertainty quantification results show relatively small biases for k-eff and reactivity. The uncertainty in peak power is around 3%, while it is negligible for the time to peak power and the pulse width. The time evolution of the standard deviation and skewness of the total power showed special shapes with relatively high maximum values. In addition, the uncertainty due to nuclear data in the two important safety parameters, i.e. maximum nodal fuel temperature and enthalpy reaches maximum value around 2% and 10%, respectively.
Journal
[en] The mechanical test procedures that address fuel cladding failure during a RIA are reviewed with an emphasis on the development of test procedures that determine the deformation and fracture behavior of cladding under conditions similar to those reached in a RIA. An analysis of cladding strain and stress are summarized and advantages and disadvantages of these tests are discussed. The short ring tests as well as EDC tests are quite different from those experienced by cladding during RIA, the burst test, the ring compression tests, the PSU plane-strain tests and the magneto-forming test can provide comparatively small corrections in simulating the stress state of RIA. However, each of these test procedures have limitations in duplicating the precise failure conditions associated with RIA. (authors)
Source
China Nuclear Society (China); 381 p; ISBN 978-7-5022-7103-9; ; Apr 2016; p. 350-357; 2015 academic annual meeting of China Nuclear Society; Mianyang (China); 21-24 Sep 2015; 6 figs., 16 refs.
; Apr 2016; p. 350-357; 2015 academic annual meeting of China Nuclear Society; Mianyang (China); 21-24 Sep 2015; 6 figs., 16 refs.
[en] This paper deals with the second phase of the RIA (reactivity initiated accident) fuel rod code benchmark for water-cooled reactors that began in 2014. This second phase has been organized around 2 complementary activities: the first activity is to compare the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena, while the second activity is focused on the assessment of the uncertainty of the results. The detailed specification and comparison of the results from the first activity is presented in this paper. Based on the Phase-I and Phase-II conclusions, some generic recommendations can be made. First, fuel and clad thermomechanical models (with the associated material properties) should be further improved and validated more extensively against a sound RIA database. Secondly, build-up of a comprehensive and robust database consisting of both separate-effect tests and integral tests should be pursued in the short term to reinforce the basis for assessment of both individual model and model integration into codes. Thirdly, an assessment of the uncertainty of fuel thermo-mechanics is of high interest (which is consistent with the second activity of this RIA benchmark Phase-II). Finally, as RIA fuel codes are more and more likely to be used for reactor accident studies, particularly for those involving safety analyses, the fuel rod failure criteria (generally used in such analyses) will have to be carefully justified and validated. The current RIA fuel failure criteria are mainly based on the fuel thermal variables and the verification is based on 'conservative' assumptions for the heat transfer conditions. As all codes give rather consistent evaluations of such variables, it appears possible, taking into account adequate provisions, to derive criteria based on thermal variables from experimental values or from an analytical approach. However, if in the future more mechanistic modelling is ever to be used to establish fuel-failure criteria based on mechanical variables, weaknesses in the validation of the codes identified above will have to be accounted for
Source
American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States); 1670 p; ISBN 978-0-89448-734-7; ; 2016; p. 219-228; Top Fuel 2016: LWR fuels with enhanced safety and performance; Boise, ID (United States); 11-15 Sep 2016; Available from: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US), also available on CD-Rom; Country of input: France; 7 refs.
; 2016; p. 219-228; Top Fuel 2016: LWR fuels with enhanced safety and performance; Boise, ID (United States); 11-15 Sep 2016; Available from: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US), also available on CD-Rom; Country of input: France; 7 refs.
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