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[en] Measurements were performed by means of the pulsed-neutron technique in a slightly subcritical, highly symmetrical configuration of the light-water-reflected, MTR-fuel-type RA-2 reactor. During measurements, the position(s) of one (two) Cd plates was (were) varied, so that the reactivity associated with each separate plate (and with both of them) was obtained. Experimental values were compared with calculated ones obtained either from simple analytical models or by the use of adequate codes. (author)
[en] The present fuel element of the ET-RR1 research reactor has a 1.75 cm lattice pitch. The neutronic studies were proved that, this lattice pitch is over moderated and not the suitable one from the fuel economic point of view. Two fuel lattice pitches are proposed, one has 1.4 cm lattice pitch with 10% U235 enrichment and the other has 1.75 cm lattice pitch with 15% U235 enrichment. The comparative neutronic study was done between these two proposed fuel lattice pitches against the present one in two cases, one for the complete core configuration of the ET-RR-1 which includes 52 fuel elements and the other for one of the actual core configuration load contains 47 fuel elements. This study is included the calculations of different neutronic parameters as the infinite and effective multiplication factor, the multi-group neutron flux along the reactor core, and the power peaking factor. The above factors were calculated by using the WIMSD4 code for lattice cell calculation, and the DIXY2 code for diffusion calculations. The results are represented in some tables and figures
[en] This study was carried out to assess the distribution of thermal neutrons emitted directly from the core of the ET-RR-1 reactor in ordinary concrete, ilmenite concrete and ilmenite-limonite concrete shields. Measurements were carried out by using a direct beam and a cadmium filtered beam of reactor neutrons. The neutron dose distributions were measured using Li2B4O7:Mn thermoluminescent dosimeters. The data obtained show that ilmenite concrete is better for slow and thermal neutron attenuation than both ordinary and ilmenite-limonite concrete. Also it was concluded that thermal neutrons emitted directly from the reactor core are highly absorbed within the first few centimeters of each type of concrete. The thickness of ilmenite concrete required to attenuate the doses of neutrons to a certain value along the beam axis for a direct reactor beam estimated to be about 75 and 57% of the shield thickness made from ordinary and ilmenite-limonite concretes, respectively. Empirical formulae were derived to calculate the neutron dose distribution in ordinary, ilmenite and ilmenite-limonite concrete shields both along and perpendicular to the beam axis for both the direct reactor neutrons and the reactor thermal neutrons. (author)
[en] Highlights: • Complete characteristics of main pump are researched into. • The quadratic character of head and torque under some operatings. • The characteristics tend to be the same under certain conditions. • The normalization method gives proper estimations on external characteristics. • The normalization method can efficiently improve the security computing. - Abstract: The paper summarizes the complete characteristics of nuclear main pumps based on experimental results and makes a detailed study, and then draws a series of important conclusions: with regard to the overall flow area, the runaway operating and 0-revolving-speed operating of nuclear main pumps both have quadratic characteristics; with regard to the infinite flow, the braking operation and the 0-revolving-speed operation show consistent external characteristics. To remedy the shortcomings of the traditional complete-characteristic expression with regards to only describing limited flow sections at specific revolving speeds, the paper proposes a normalization method. As an important boundary condition of the security computing of unstable transient process of the primary reactor coolant pump and the nuclear island primary circuit and secondary circuit, the precision of complete-characteristic data and curve impacts the precision of security computing. A normalization curve obtained by applying the normalization method to process complete-characteristic data could correctly, completely and precisely express the complete characteristics of the primary reactor coolant pump under any rotational speed and full flow, and is capable of giving proper estimations on external characteristics of the flow outside the test range and even of the infinite flow. These advantages are of great significance for the improvement of security computing of transient processes of the primary reactor coolant pump and the circuit system.
[en] A 3D neutronic model for the RA-3 reactor was developed on the basis of previous experience and validated with selected experimental data. Control rod calibrations were reproduced in N94 and N136 cores. The calculated values are shown to be dependent on relative position of the rods and the procedure that gives the best estimation of the rod value is the one performed following the experimental method compensating small rod insertions with small extractions. Rod worth calculations differ from the measured values in less than 2%. Rod-drop experiments were used to evaluate rod effectivities. The experimental results showed discrepancies between estimators derived from the point reactor model, and from spatial modal kinetics. Discrepancies are also observed when using different detectors. Even when using the spatial modal kinetics approach, the estimators obtained from different detectors disagree when one of them is located near to the rod, but differences are considerably reduced with respect to point reactor model because in this case only the delayed evolution is considered. We can say that all estimators give fairly similar results when the detector field of view is not influenced by the local perturbation introduced by the falling rod. This indicates the existence of spatial effects which are not completely accounted for in the spatial modal kinetics approach. Also, the importance of verifying the form function behaviour during the delayed evolution. The rod drop experiments were simulated using the improved quasi-static model and static evaluations. The rate of overestimation static/dynamic is constant in both core configurations and varies between 18% and 23% for the analyzed rods. The dynamic model allows comparing also thermal flux ratios at the detector positions. The neutronic model is considered reliable for design and fuel management analysis. The estimations of criticality, control rod calibrations and excess reactivity are satisfactory and simple models representing the core components without any kind of correction factors have been used to achieve these results
[en] An automated software, BMAC, for modeling and performing the neutronics calculations of MNSRs and similar reactors (TRIGAs) has been developed. Calculation of initial excess reactivity, flux and power distributions, and all other neutronic parameters of the reactor, full core representation, can be made automatically using a 3-D model, by coupling WIMSD-4 and CITATION codes, in a very quick and simple way. No preliminary CITATION input file is needed. All required data are read from an external input file simply prepared. Accurate results for the parameters of the reactor, in the framework of Diffusion Theory, can be obtained
[en] A pulsed neutron polyenergetic thermal beam at ET-RR-1 is produced by a phased double-rotor facility. One of the rotors has two diametrically opposite curved slots, while the second is designed to operate as a rotating collimator. The dimensions of the phased rotating collimator are selected to match the curved slot rotor. The calculated collimator transmissions at different operating conditions are found to be in good agreement with the experimental ones. The optimum operating conditions of the double-rotor facility are deduced. The calculations were carried out using a computer program RCOL. The RCOL was designed in FORTRAN-77 to operate on PCs. (author)
[en] Monte Carlo modeling of the Kalpakkam Mini Reactor (KAMINI) has been carried out for the first time by using Monte Carlo code (MCNP4A) and continuous energy cross-sections. The safety control plate (SCP) drop experiment is simulated and the computed integral worth of the SCPs is compared with the measured value. The measured axial neutron flux profile and foil reaction rates in one of the in-core irradiation location and the foil reaction rates at the west beam port are also compared with the predicted results. The agreement between measurements and calculations is quite satisfactory. It is confirmed from the calculation and measurement that the north thimble is having nearly 10-20% higher neutron flux as compared to the south thimble depending on the exact elevation
[en] The power excursion characteristics of the Nigeria Research Reactor-1 (NIRR-1) under different reactivity insertion transients have been calculated using PARET/ANL 7.3 code. Experimental data of dynamic experiments performed in NIRR-1 facility during initial startup were used to benchmark the calculated data. For ramp insertion of total cold core excess reactivity of 3.77 mK, the reactor quickly reaches a maximum power of 82.3 kW in 390 s and decreases gradually due to the inherent safety features. The calculated maximum power for this insertion is in good agreement with a measured value of 79.4 kW recorded during the commissioning of NIRR-1. Similarly, moderator temperature for the hottest channel with respect to this transient was calculated to be 63.7 deg. C which is comparable to a measured value of 63.9 deg. C. Furthermore, step reactivity insertions of 2.23 mK and an approximately 0.32 mK, a regular criticality experimental check carried out at MNSR facilities by Safeguards Inspectors were accurately simulated.
[en] Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification