#Iaea flux tecdoc code#

The value of the average thermal flux in the central trap for the BOL case is 2.255E+14, which was calculated by the ANL codes as 2.585E+14 The thermal flux slightly increases at the interface of the fuel/water and fuel/graphite.

The effective delayed-neutron fraction was maximum at 0.007185 for the original aluminum clad while it was minimum at 0.007078 for stainless steel clad. maximum thermal flux occurs at the central flux trap and drops exponentially, moving away from the core. The prompt neutron generation time was maximum at 45.36В Ојs when stainless steel-316 was the clad while it was minimum at 44.03В Ојs for the original aluminum clad. The thermal neutron flux at the mid of the central flux trap increased by 59.9% and 12.5% for stainless steel and zircaloy-4 clads, respectively, from the flux of the original aluminum clad. It was observed that at the beginning of life, the excess reactivity was maximum at 0.054110 О” k/ k when zircaloy-4 was used as clad while it was minimum at в€’0.365650 О” k/ k when stainless steel-316 was the clad as compared to 0.017945 О” k/ k for aluminum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. The reactor analyzed is the same one utilized for the benchmark problem solved in IAEA-TECDOC-233 (1980), with the water in the central flux trap replaced with a 7.7 cm Г— 8.1 cm block of aluminum containing a square hole 5.0 cm on each side (IAEA-TECDOC-643, 1992) ().Description of the low enriched uranium core of the reactor is given in Table 1. It presents the main applications utilizing low and medium flux neutron sources, that could be the object of research and development or are relevant in education and training.

Calculations were carried out to find the core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed-neutron fraction. An overview of up-to-date technical information on neutron scattering techniques and on instrumentation for neutron scattering is presented in this publication, which draws upon the outcome of an IAEA technical meeting. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Finally, some implications of this method are mentioned regarding future work required in order to exploit fully its potential.The kinetic parameters of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. From this discussion it is concluded that the proposed least-squares method does provide the best complete solution, with uncertainties, to the problem as it is understood now. It is shown that the least-squares method does not require more input information than would be needed by our current methods in order to estimate the uncertainties in their solutions. The relationship of the solution to this least-squares problem to those obtained currently by other methods of solving the dosimetry neutron spectrum unfolding problem is extensively discussed.

#Iaea flux tecdoc code#

Two interpretations are made for the solution of the code STAY'SL, which solves this least-squares problem. Therefore, the solution should not be obtained by iteration. The fact that the problem is technically nonlinear, but should be treated in general as a linear one, is argued. A chi 2-test is derived for the consistency of the input data, which does not require the solution to be first obtained. The existence of this solution is analyzed in some detail. The mathematical solution to this least-squares problem is derived from the general solution. From this analysis, a least-squares approach to the solution of the dosimetry neutron spectrum unfolding problem is introduced.

To achieve this, one has to optimize the neutron flux and neutron energy. The method of least-squares is briefly reviewed, and the conditions under which it may be used are stated. IAEA-TECDOC-1516 Viability of inert matrix fuel in reducing plutonium amounts.