VERIFICATION OF SAPFIR_95&RC_VVER PACKAGE FOR CORE NEUTRONIC CALCULATIONS IN EXTENDED RANGE OF PERMISSIBLE VALUES OF VVER PARAMETERS
24th Symposium of AER on VVER Reactor Physics and Reactor Safety (2014, Sochi, Russia)
Reactor physics calculations, experiments and code validation
Abstract
VERIFICATION OF SAPFIR_95&RC_VVER PACKAGE FOR CORE NEUTRONIC CALCULATIONS IN EXTENDED RANGE OF PERMISSIBLE VALUES OF VVER PARAMETERS
Artemov V.G., Artemova L.M., Zinatullin R.E., Ivanov A.S., Kuznetsov A.N., Piskarev A.V., Shemaev Yu. P.,
FSUE «Alexandrov NITI», Sosnovy Bor, Russia
Kurakin K.Yu, Tihomirov A.N., Fateyev M.V. OKB «. Gidropress», Podolsk, Russia
ABSTRACT
The SAPFIR_95&RC_VVER package is designed for calculation of VVER neutronic characteristics, critical fuel assemblies, and computational polygrids of fuel storage pools.
The SAPFIR_95&RC_VVER package was certified by the Nuclear and Radiation Safety Center in 2005. It is now used in OKB “Gidropress” for both safety analysis of fuel loaded in operating reactors and design of advanced reactor cores.
In view of required increase in power of the operating VVERs and their longer operation period between refuelings, the design of the VVER fuel elements and assemblies was modified. An additional verification of the SAPFIR_95&RC_VVER package has been carried out to demonstrate that it is applicable to a safety analysis of cores with modified fuel taking into account changes in permissible values of fuel assembly and core parameters.
The additional verification has shown that neutronic calculation errors for the cores with modified fuel assemblies agree with errors given in the SAPFIR_95&RC_VVER Certificate 2005. Also, the verification has established lower calculation errors for critical boron concentration, efficiency of control rods, and worth of shutdown rods than those given in Certificate 2005.
The paper presents results of verification calculations on key neutronic parameters. The verification used experimental data on a series of fuel batches representing sequential introduction of new type fuel assemblies in the operating power units of the Balakovo NPP and the Rostov NPP. The neutronic calculation error for coolant boiling conditions is verified against experiments performed in the operating ВК-50 (BWR type) reactor. The calculation accuracy of power in individual fuel rods is verified against results of benchmark Monte- Carlo calculations.