Accuracy analysis of BN-600 power density calculations

Abstract

Methodological program techniques for performing neutron-physical calculations of fast-reactor cores with a description of the main codes TRIGEX, JARFR, GEFEST, MMKKENO, and ModExSys are described. The results of verification and assessment of the methodological errors on the basis of a test model of BN-600 are presented. Transport effects and mesh errors in the reaction rate distributions in the core, lateral screen, and in-reactor storage area are evaluated. An integral assessment of the computational accuracy is performed by comparing with the experimental data obtained by γ-scans of BN-600 fuel assemblies. It is shown that calculations describe the experimental data over the core with an error no worse than 5%. In the lateral screen and the in-reactor storage area, the discrepancy between the calculations and experiment does not exceed 20-30%.Computational modeling of the neutron field is the main method of determining the energy release in a fast-reactor core, where, as a rule, there is no apparatus for tracking the power density. Correspondingly, computational accuracy is one of the key elements of the operational reliability of fuel elements and assemblies. Another aspect of the problem is substantiating the conservatism of fast-reactor core design -design limits or margins affecting the technical-economic characteristics. Experience in operating BN-600 makes it possible to evaluate the accuracy of the design and operational programs under real conditions of a commercial medium-power fast reactor and to take it into account in subsequent development work.Such an analysis is based on the measurement of the power density in BN-600 by means of γ-scanning. Eleven such experiments were completed in the time period from physical startup of the reactor in 1980 and up to 2008. The present article analyzes the last experiments performed in 2003-2006 in the course of transitioning from a 01M1 core to a 01M2 core with maximum burnup 11.1% [1]. This upgrade was an important next step in making use of the serviceability margin of reactor fuel assemblies, which required greater attention to monitoring the characteristics of the core [2].Program and Constants System for Neutron-Physical Calculations of Fast Reactors. A consistent system for performing neutron-physical calculations of fast reactors is now available. It includes the program systems TRIGEX [3], JARFR [4], GEFEST [5], and MMKKENO [6] as well as the BNAB constants library [7,8]. It should be noted that the base

Precision neutronic calculations of experiments on the neutron transmission through the reflector layers at the BFS critical facilities for expanding the verification database to justify lead cooled fast reactor designs

“…The calculation and experiment discrepancies in the distributions of the 235 U and 239 Pu fission reaction rates in the steel radial blanket of a fast reactor at a level of 20 to 25% were recorded earlier in (Danilychev et al 2002;Kochetkov et al 2007;Mitenkova and Novikov 2011a). The results obtained in the study do not contradict to the calculation accuracies for the distributions of the power density field in a steel blanket evaluated earlier in (Blyskavka et al 2010;Mitenkova and Novikov 2011). An analysis of the BFS-66-B series experiments has however shown that a reduction in the calculation and experiment discrepancies to the measurement error level for the Cr 18% and Cr 28% steel grades can be explained by an overcompensati- on of errors in determining the Fe, Cr and Ni isotope neutron cross-sections.…”

Calculation analysis of experiments on passing neutrons through the reflector layers at the BFSs critical assemblies for complementing the lead-cooled fast reactor verification database

“…Расчетно экспериментальные расхождения распределений скоростей реакций деле ния 235 U и 239 Pu в стальном боковом экране быстрого реактора, лежащие на уровне 20 -25%, отмечались ранее в [10 -12]. Полученные в работе результаты не противоречат ранее оцененным в [13] точностям расчета распределения поля энерговыделения в стальном экране. Однако анализ экспериментов серии БФС 66 Б показал, что снижение расчетно экспериментальных расхождений до уровня погрешности измерений для ста лей марки X18 и X28 может быть обусловлено перекомпенсацией ошибок в определе нии сечений изотопов Fe, Cr и Ni.…”

Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries

“…was calculated using a Monte Carlo code(Blyskavka et al 2001, Zherdev et al 2018 with the BNAB-93 and BNAB-RF group neutron constants, as well as with evaluated neutron data files of the Russian ROSFOND library and versions of U.S., European and Japanese libraries from different years (ENDF, JEFF, JENDL).…”