Dynamic simulator for the miniature neutron source reactor

A dynamic simulator for the Syrian Miniature Neutron Source Reactor was developed and implemented on a desktop computer using C++ Builder. Mathematical models for the main physical phenomena of reactor such as heat transfer and neutronics were developed on the basis of the lumped parameter approach and real experimental data fitting. Point model equations of reactor kinetics was employed and solved using fourth order Runge-Kutta integration procedure.

Simulation for training purposes of both real and accelerated time for normal and abnormal conditions can be accomplished with the model. The simulator is user friendly with operator.     

The effect of temperature and control rod position on the spatial neutron flux distribution in the Syrian miniature neutron source reactor

The effect on the spatial neutron flux distribution for both of water and fuel temperature increase as well as the change in the control rod position are presented in the Syrian miniature neutron source reactor (MNSR). The cross-sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the spatial neutron flux distribution at different water and fuel temperatures and different control rod positions using four energy groups. This work shows that the increase in water and fuel temperatures during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. The change in the control rod position does not affect as well the spatial neutron flux distribution in the reactor except in the region around the control rod position.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C

Based on probabilistic approach, the MCNP-4C code has been used effectively to simulate the Syrian MNSR reactor core and all its surrounding components in three dimensions, including a preliminary conceptual design of a thermal column to be installed later. For verification and validation purposes, reactor calculations include: criticality and control rod worth. Values of these parameters are 1.00517 and 6.54 mk, respectively. The thermal column is to be installed in the water of the reactor pool. Optimal conditions for this thermal column were tested using the already developed model. Optimization focused on the most suitable position for placement of the column in the water pool, dimensions, and material. The aim was to have a thermal neutron flux of 1 × 10⁹ n cm⁻² s⁻¹ in the center of thermal column, and resonant and fast neutron fluxes to be as low as possible as well.     

Simulation of reactivity transients in a miniature neutron source reactor core

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease in value due to temperature feedback effects. The trends of theoretical results were found to be similar to measured values and the peak powers agreed well with experimental results. For ramp reactivity equivalent of clean core cold excess reactivity of 4 mk (4×10⁻³ Δk/k), the predicted peak power of 100.8 kW agrees favourably with the experimental value of 100.2 kW. The measured outlet temperature of 72.6 °C is also in agreement with the calculated value of 72.9 °C for the release of the core excess reactivity. Theoretical results for the postulated accidents due to fresh fuel replacement of reactivity worth 6.71 mk and addition of incorrect thickness of Be plates resulting in 9 mk reactivity insertion were 187.23 and 254.3 kW, respectively. For these high peak powers associated with these reactivity insertions, it is expected that nucleate boiling will occur within the flow channels of the reactor core.     

Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactors

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.     

Xenon poisoning calculation code for miniature neutron source reactor（MNSR）

In line with the actual requirents and based upon the specific characteristics of MNSR,a revised point-reactor model was adopted to model MNSR‘s xenon poisoning.The corresponding calculation code.MNSRXPCC(Xenon Poisoning Calculation Code for MNSR),was developed and tested by the Shanghai MNSR data.     

Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step- ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the Safety Analysis Report.     

A thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients

The transient thermal-hydraulic problem of MNSR is represented by ten differential equations solved numerically using Runge–Kutta method.Computational results are then compared with experimental measurements. Fuel grids and cooling coil models are incorporated in the model too. Radiating energy from the clad is taken into account in the energy balance in the reactor. The pool is divided into three sections in the model. The effect of the cooling coil of the pool upper section on reactor thermal-hydraulic parameters is discussed. The only input parameter of the reactor is the power temporal distribution. Good agreement between calculated and measured data was obtained.     

Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core

The total neutron flux spectrum of the compact core of Ghana’s miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) × 10¹² n/cm² s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 10¹¹ n/cm² s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) × 10¹¹ n/cm² s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) × 10⁻⁰⁸ MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) × 10⁰⁹ n/cm² s at the lower energy end of the slowing down region between 8.2491 × 10⁻⁰¹ MeV and 8.2680 × 10⁻⁰¹ MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) × 10⁰⁸ n/cm² s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.     

Dynamic analysis of the closed-loop transfer function in the miniature neutron source reactor (MNSR)

The closed-loop transfer function of Syrian miniature neutron source reactor (MNSR) has been measured experimentally using the prompt jump approximation technique. Analysing the reactor stability behaviour, a physical model has been formulated based on the open-loop (neutronics) transfer function employing the lumped parameter concept to describe the reactor thermohydraulic characteristics. The reactor kinetics is described by the point kinetic model for one-group of delayed neutrons. Inherent internal feedback effect is considered as a single reactivity feedback that represents the coolant temperature effect. Comparison of the analytically derived transfer-function with the experimental one shows good agreement. Stability analysis of the closed-loop transfer function has been made using the Nyquist criterion and Bode diagram. Routh–Hurwitz criterion has been applied to estimate the stability limit of the MNSR closed-loop. The Nyquist and Bode criteria have shown that the MNSR closed-loop transfer function is indeed stable. The Routh–Hurwitz criterion enabled the estimation of the upper limit of temperature feedback coefficient of reactivity. Results indicate that MNSR has high inherently safety features. Various relationships that govern relation amongst reactor variables such as the isothermal reactivity coefficient of moderator temperature, temperature difference across the core and coolant flow rate of the natural circulation and mean time for heat transfer to the coolant have been concluded.     

On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.     

The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor

In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (T_in) and the temperature difference between the coolant inlet and outlet (ΔT) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong dependence of the reactor power on coolant temperature in agreement with the design of MNSR. The experimental parameters were found to be in good agreement with data obtained using a semi-empirical relationship between the reactor power, flux parameters, core inlet temperature, and the coolant temperature rise. The relationship was therefore used to predict the power level of NIRR-1 from its neutron flux parameters to which it has been found to be proportional. The variation of T_in and ΔT with the reactor power and flux was also investigated and the results obtained are hereby discussed.     

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well.     

反应堆冷中子源中子物理学计算

用MCNP软件计算反应堆冷中子源,慢化剂室内平均中子注量率为6.69× 1013/cm-2.s-1,波长为0.4 nm和0.6 nm的冷中子增益因子～16和32.冷源慢化剂中正仲氢比例对输出的冷中子能谱有较大影响,而在3K范围内慢化剂温度变化对冷中子能谱的影响很小.计算结果表明,冷中子源性能达到基本设计要求.     

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF₃ were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.     

Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor

We appraised in this study the effects of core excess reactivity and average coolant temperature on the operable time of the Nigeria Research Reactor-1 (NIRR-1), which is a miniature neutron source reactor (MNSR). The duration of the reactor operating time and fluence depletion under different operation mode as well as change in core excess reactivity with temperature coefficient was investigated over a period of five years. Our result shows that there is a strong dependence of reactor operating time on core excess reactivity and temperature coefficient. It was observed in 2004 that with a cold core excess reactivity of 3.77 mk, at full-power flux of 1.0 × 10¹² n cm⁻² s⁻¹ the reactor operated for 5 continues hours. At half-power flux of 0.5 × 10¹² n cm⁻² s⁻¹ and under the same excess reactivity condition, the reactor reaches 8 h of operation. However, re-measurements done in 2009 shows that excess reactivity of the reactor has reduced to 2.80 mk, the operable time at full flux dropped to 3.5 h while that of half-power became 7 h. We also investigated the reactor's energy consumption within the period under study and found to be much more in 2008 compared to the previous years. We infer that the amount of fluence consumed and the excessive reactor usage in 2008 has contributed significantly to the reduction of the reactor's excess reactivity in that year. The results obtained here revels that for an MNSR with a clean core excess reactivity between 3.5 mk and 4.0 mk, 5 and 8 h are the maximum operable times under full and half-power flux conditions, respectively. Negative deviation from these optimum times is therefore an indication of a drop in excess reactivity and the need for beryllium shims addition.     

Flow excursion time scales in the advanced neutron source reactor

Flow excursion transients give rise to a key thermal limit for the proposed advanced neutron source (ANS) reactor because its core involves many parallel flow channels with a common pressure drop. Since one can envision certain accident scenarios in which the thermal limits set by flow excursion correlations might be exceeded for brief intervals, a key objective is to determine how long a flow excursion would take to bring about a system failure that could lead to fuel damage. The anticipated time scale for flow excursions has been examined by subdividing the process into its component phenomena: bubble formation, flow deceleration, and fuel plate heat-up. Models were developed to estimate the time required for each individual stage. Accident scenarios involving sudden reduction in core flow or core exit pressure have been examined, and the models compared with RELAP5 output for the ANS geometry. For a high-performance reactor such as the ANS, flow excursion time scales were predicted to be in the millisecond range, so that even very brief transients might lead to fuel damage. These results have been useful for determining the significance of momentary flow excursion events calculated for accident situations in the ANS reactor. In addition, the methods presented are applicable for evaluating the timing of flow excursion transients in other facilities as well.