Development, Validation and Applications of SRAC: JAERI Thermal Reactor Standard Neutronics Code System

A study on the anisotropic scattering effects in heterogeneous square cells of light water reactors has been performed using the characteristics method. It was found that the effects of the anisotropic scattering were relatively large for the MOX fuel cell because of the large neutron current from the moderator to the fuel region and the k inf value by the P 0 calculation became 0.10-0.16% larger than that by the P 5 calculation. With the transport correction, the k inf difference from the P 5 calculation became even larger than that from the P 0 calculation and the k inf value by the transport correction became 0.18-0.25% larger than that by the P 5 calculation for the MOX fuel cell. The transport corrected self-scattering cross sections of the moderator region become smaller than the non-transport corrected ones and the angular flux distribution becomes more anisotropic with the transport correction. Therefore, more neutrons toward the moderator region between the fuel pellets can slow down to the lower energy region with the transport correction. As a result, the k inf value by the transport correction becomes larger than that by the P 0 calculation, which is opposite effect to that by the P 5 calculation.

Section: Numerical Results Of the Anisotropic Scattering Effect Imentioning

confidence: 99%

Neutron Anisotropic Scattering Effect in Heterogeneous Cell Calculations of Light Water Reactors

Ushio

Takeda

Mori³

2003

“…This is partly because the a (=a cl a 1 ) value is larger in the higher energy range in thermal energy and partly because the contribution of 239 Pu, whose a is larger than that of 235 U, increases. The reason why kin£ increases by 2% in spite of the small increase of fission is that the v value of 239 Pu is larger than that of 235 U.…”

Section: Coolant Void Effect On Reaction Ratementioning

confidence: 93%

“…A series of cell calculations was carried out using the SRAC code system< 1 > to provide the fundamental neutronic characteristics to the succeeding accident analysis.…”

Section: Introductionmentioning

confidence: 99%

Analysis of Reactivity Coefficients of Chernobyl Reactor by Cell Calculation

Tsuchihashi¹,

Akino²

1987

Self Cite

A series of cell calculations for the Chernobyl reactor was performed using the SRAC code system to provide its fundamental neutronic characteristics for the accident analysis at JAERI. The calculations are based on a two-step cell modelling. The primary cell is supposed on a unit square graphite block of 25 em x 25 em which contains a fuel assembly or a control channel. The secondary cell is supposed on a unit of 16 channels where 14 fuel and 2 control channels are located so as to simulate the whole core.Detailed investigation on fractional change of reaction rates for each nuclide along with increase of void fraction was carried out. The analysis clarified the mechanism to induce the positive void coefficient, together with its burn-up dependence and the increase due to withdrawal of control rod.A comparison of the effect of void fraction on the reaction rates in the primary cell between a Monte Carlo code VIM and SRAC shows consistent results within the statistical error.The calculated results for the composition of discharged fuel, void and other reactivity coefficients, kinetic parameters and their burn-up dependence show satisfactorily good agreement with those reported by the Soviet Union and some institutions.

“…Since the functions have been qualified through a wide range of benchmark calculations including very high temperature reactor critical assembly (VHTRC) [17,18] and HTTR [19], the SRAC code system has been used for neutronics design of the core conversion of JAEA research and test reactors and the analyses of critical experiments performed at JAEA and domestic universities [20].…”

Section: Reactor Kinetics Modelmentioning

confidence: 99%

Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

Takamatsu

Tochio

Nakagawa

et al. 2014

In a safety demonstration test involving the loss of both reactor reactivity control and core cooling, the high-temperature engineering test reactor (HTTR) demonstrates spontaneous stabilization of the reactor power. The test and analytical results of tripping one or two out of three gas circulators without reactor scram have already been reported. Moreover, the pre-analytical result of tripping all three gas circulators without reactor scram has been presented. On the other hand, the test and analytical results of tripping all three gas circulators without reactor scram are shown in this paper. About experiments, at an initial reactor power of 30% (9 MW), when all three gas circulators were tripped without reactor scram to reduce the coolant flow rate to zero, the fuel temperature did not show a large increase because the large heat capacity of the graphite core could absorb heat from the fuel in a short period. Moreover, the decay heat could be transferred through the graphite core and the reactor pressure vessel (RPV), emitted by thermal radiation from its outer surface and removed to the active vessel cooling system; therefore, the core at 9 MW was never exposed to the danger of a core melt, and the reactor power was stabilized spontaneously. About analyses, the reactivity performance is important for predicting the converging level of reactor power that affects the fuel temperature during a loss of forced cooling (LOFC) without reactor scram. With regard to thermal hydraulics, the performances of graphite heat conduction in the reactor core and thermal radiation from the RPV surface to the reactor cavity cooling system are crucial for predicting the temperature behavior of the fuel and RPV in the LOFC condition. It was confirmed that reactor kinetics coupled with heat transfer could be applied to reactor safety and accident analysis based on the comparison between the experiments and the analyses.