Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.     

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.     

Prospects of using different clad materials in a material test research reactor – Part 1 – The kinetic parameters

The kinetic parameters of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Calculations were carried out to find the core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed-neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that at the beginning of life, the excess reactivity was maximum at 0.054110 Δk/k when zircaloy-4 was used as clad while it was minimum at ?0.365650 Δk/k when stainless steel-316 was the clad as compared to 0.017945 Δk/k for aluminum. 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. The prompt neutron generation time was maximum at 45.36 μs when stainless steel-316 was the clad while it was minimum at 44.03 μs for the original aluminum clad. 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.     

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction (β_eff) and the prompt neutron generation time (Λ). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both β_eff and Λ.     

Flow of kinetic parameters in a typical swimming pool type research reactor

Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The β_eff decreases with 1.33 × 10⁻⁶/% burnup whereas prompt neutron generation time increases with 6.42 × 10⁻⁸ s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core.     

Study of kinetic parameters in the RA-4 reactor by means of computational modeling and neutron noise measurements

It was carried out a study of the kinetic parameters of the reactor RA-4 (SUR-100 type reactor) by means of a simulation with calculation codes and a measurement with the neutron noise technique. For the simulation the WIMS (cell calculation) and PUMA (3D diffusion code to model a complete reactor) codes were used to estimate the ‘effective delayed neutron fraction’ (β_eff), the ‘prompt neutron generation time’ (Λ), and the ‘prompt neutron decay constant’ (α). In addition, a sensitivity analysis was made to determine the best nuclear data evaluations for this model and the influence of variables such as the mean speed of neutrons, the nuclear delayed neutron fraction and the delayed neutron fission spectrum. The neutron noise technique was used to determine α. The comparison between the results for the value of α obtained by the simulation and the experience shows a good agreement. In this way the validity of a diffusion calculation for a small reactor with important heterogeneities is verified, making certain modifications to avoid the limitations of this theory.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique

A comprehensive 3-D model of the Syrian MNSR reactor has been developed using the MCNP-4C code aiming at accurate predicting of key core physics parameters. For the currently utilized HEU fuel (89.87% UAl4-Al) and two possible alternative LEU fuels (UO2 12%, and UO2 20%) the main core kinetics parameters like prompt neutron generation time, effective delayed neutron fraction, clean cold core excess reactivity and reactivity feedback coefficients of moderator temperature have been calculated. In this regard the role of particle weight loss on capture, fission and escape in determining the temperature effect of reactivity has been evaluated. The calculated results for the HEU fuel agree well with experimental values. The evaluated kinetics parameters are being used in accomplishing necessarily safety analyses related to the conversion of MNSR reactor to low enriched uranium.     

Calculation of the effective delayed neutron fraction and the prompt neutron lifetime with the indigenously modified CEBIS code: MCEBIS

The CEBIS code has been modified to enable the calcination of both the effective delayed neutron fraction and prompt neutron generation lifetime in any nuclear thermal reactor, especially reactor types such as TRIGA, SLOWPOKE, and MNSR. The new version, called MCEBIS, includes sonie special subroutines which will be called up as part of the input to calculate the above two dynamic parameters. In addition, some control flags have been added to recognize any important reactor components such as beryllium as a reflector or heavy water as moderator and pence calculate their photo-neutron fractions.

The MCEBIS code has been tested using two reactor models: TRIGA and MNSR. These models were developed mainly to verify the modified code. Each model represents a 1-D neutronics model of the reactor. Calculated results for the effective delayed neutron fraction and prompt neutron generation lifetime in both reactors have been compared with published data. Good agreement with published results has been established.     

基于OpenMC的瞬发中子衰减常数计算模块开发与验证

瞬发中子基波衰减常数α可定量描述反应堆内中子随时间的变化,是计算绝对反应性所需的中子动力学参数之一,对次临界（特别是较深次临界）绝对反应性的精确测量具有重要意义。本文在开源程序OpenMC基础上,基于k α迭代方法,以中子径迹长度上的平均时间吸收权重修正作为k α迭代参数因子,在输运过程中对瞬发、缓发中子分别考虑,开发了具有瞬发α本征值问题计算功能的OpenMC PA模块。以Godiva衍生基准题和MUSE 4次临界实验装置为计算对象,对程序计算瞬发α本征值问题能力进行验证。结果表明,该计算模块有优于MCNP4C的计算速度与计算范围,计算值与参考值的相对误差小于05%。OpenMC PA能满足次临界系统瞬发α本征值和中子动力学参数计算需求。     

Monte Carlo simulation of Feynman-α and Rossi-α techniques for calculation of kinetic parameters of Tehran Research Reactor

Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α₀), effective delayed neutron fraction (β_eff) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1 W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are validated against both the experimental data and the results of MTR_PC computer code.     

Development of an improved method for calculating the effective delayed neutron fraction of a core using a structure with only a few collapsed energy groups

In this study, an improved method for calculating the effective delayed neutron fraction of a core with a few collapsed energy groups has been developed. To accurately calculate the effective delayed neutron fraction of a core using the conventional method, a structure with several energy groups is needed for the fast energy region in order to reflect the difference in the fission spectra for prompt fission neutrons and delayed neutrons. On the other hand, a structure with only a few energy groups is needed for the criticality evaluation. Thus, the calculation cost increases for the effective delayed neutron fraction calculations owing to the need for a large number of energy groups. To solve this problem, in the present study, the error mechanism for the effective delayed neutron fraction calculation using a structure with only a few energy groups was studied, and it was found that the error results from the collapse of the fission spectra after the cell calculations without adjoint flux weighting. In addition, an improved method for the collapse fission spectra with an adjoint flux obtained by one-point calculation was developed. Using the proposed method, the effective delayed neutron fraction can be estimated with sufficient accuracy using a structure consisting of only a few collapsed energy groups. This result will contribute to reducing the calculation cost and/or improving the accuracy of effective delayed neutron fraction calculations.     

300#反应堆瞬发中子衰减常数测量

利用反应堆噪声分析技术测量300^#池式研究堆缓发临界下的瞬发中子衰减常数。堆芯采用低富集度U燃料装载，燃料元件带一定燃耗。利用紧靠堆芯布置的两个中子探测器，信号经测量系统和相关软件得到互谱密度，用非线性最小二乘法拟合得到瞬发中子衰减常数。在4kW功率水平测得缓发临界下的瞬发中子衰减常数α_c=(83.4±0.7)s^.-1。     

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.     

Evaluation of the fraction of delayed photoneutrons for TMSR-SF1

The 10 MW_(th) solid-fueled thorium molten salt reactor(TMSR-SF1) is a FLi Be salt-cooled pebble bed reactor to be deployed in 5–10 years, designed by the TMSR group. Due to a large amount of beryllium in the core, the photoneutrons are produced via(γ , n) reactions.Some of them are generated a long time after the fission event and therefore are considered as delayed neutrons. In this paper, we redefine the effective delayed neutrons into two fractions: the delayed fission neutron fraction and the delayed photoneutron fraction. With some reasonable assumptions, the inner product method and the k-ratio method are adopted for studying the effective delayed photoneutron fraction. In the k-ratio method, the Monte Carlo code MCNP6 is used to evaluate the effective photoneutron fraction as the ratio between the multiplication factors with and without contribution of the delayed neutrons and photoneutrons. In the inner product method, with the Monte Carlo and deterministic codes together, we use the adjoint neutron flux as a weighting function for the neutrons and photoneutrons generated in the core. Results of the two methods agree well with each other, but the k-ratio method requires much more computing time for the same precision.     

瞬发和缓发γ射线对堆内构件释热率影响的研究

为提高核电设计中反应堆堆内构件释热率计算的准确性,本文在原来MCNP外中子源模型计算方法的基础上,计算分析瞬发裂变γ对堆内构件释热率的贡献。计算结果显示,考虑瞬发裂变γ使得堆内构件的释热率增加9%～38%,离堆芯越近的堆内构件的增加值越大。另外,分析认为缓发γ对堆内构件释热率的贡献与瞬发裂变γ相当。因而反应堆堆内构件释热率计算中除了考虑中子及中子俘获所生γ的贡献,还应该考虑瞬发裂变γ和缓发γ的贡献。     

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.     

Basic consideration of a nuclear power monitoring system using neutron-induced prompt gamma rays

ABSTRACT

Monitoring only neutron flux in a nuclear reactor core has an advantage of offering reactor power monitoring accuracy. We started the development of a new nuclear instrumentation based on the measurement of prompt gamma rays emitted from metals placed at the neutron flux monitoring positions. The thermal neutron flux at the position of each placed metal piece can be monitored by measuring the prompt gamma rays as the count rate of each energy. The gamma-ray energy range was limited from 6 to 10 MeV to mitigate the interference of environmental gamma rays. Four metals, Ti, V, Ni, and Cu, were chosen as candidates in consideration of their neutron emission rates and self-absorption. In an experiment with a high-purity germanium semiconductor detector, we considered the identification of individual peak energies in an assumed situation where prompt gamma rays were emitted from the four different metals at the same time. Energy resolutions of the peak with the largest energy gap from the nearest energy peak of the other candidate metals were smaller than the gap. Thus, we confirmed that at least one peak for each candidate metal was able to be separated from the peaks derived from other candidate metals.     

Method of experimental estimation of the effective delayed neutron fraction and of the neutron generation lifetime in the IBR-2 pulsed reactor

Experimental investigations of random fluctuations in the power of the IBR-2 periodic pulsed reactor carried out in the steady-state mode of its operation and in the pulsed mode made it possible to obtain estimations of the main parameters of the kinetics – the effective fraction of delayed neutrons and the average lifetime of prompt neutrons. Functionals were measured relating the main parameters of the kinetic behaviour: the prompt neutron decay constant (steady-state mode) and the relative dispersion of pulse energy fluctuations. It was shown that the experimental values of parameters of the kinetic behaviour are close to the calculated ones. In addition, the power of the spontaneous neutron source was estimated on the basis of the analysis of neutron noise.