Determination of Tehran Research Reactor power by N gamma detection

The radioisotope ¹⁶N is produced by the interaction of fast neutrons with ¹⁶O in water reactor coolant. This radioisotope emits at the two major gamma ray energies of 6.13 MeV and 7.1 MeV. Exploiting the linear relation between the number of gamma particles versus the reactor power change, the reactor power is determined by detecting and counting the emitted gammas. In this work, for the detection of gammas to measure the reactor power, two different methods are employed. First, by NaI(Tl) scintillator detector and second, by assembly of ten GM detectors. The obtained results confirm that the number of emitted gammas is proportional to the change in reactor power as shown by different monitoring systems such as UIC, CIC, FC, Cherenkov and thermal power. Both of the applied methods are shown to give reliable results for reactor power above 20 kW. Both systems, having been calibrated, are being used as monitoring systems of power in Tehran Research Reactor. These systems are usable in other research reactors and possibly in power reactors as well.     

Reactor core power measurement using Cherenkov radiation and its application in Tehran Research Reactor

Cherenkov radiation is a process that could be used as an extra channel for power measurement to enhance redundancy and diversity of a reactor. This is especially easy to establish in a pool type research reactor. A simple photo diode array is used in Tehran Research Reactor to measure and display power in parallel with the existing conventional detectors. Experimental measurements on this channel showed that a good linearity exists above 100 kW range. The system has been in use for more than a year and has shown reliability and precision. Nevertheless, the system is subject to further modifications, in particular for application to lower power ranges.     

Neutronic analysis for Tehran Research Reactor mixed-core

In this research, neutronic calculation of current low enriched uranium control fuel elements replacement with high enriched uranium control fuel elements in the reference core of Tehran Research Reactor (TRR) has been investigated and the results of calculations are compared with the TRR neutronic safety criteria. Results show that all neutronic parameters of the reference and each mixed-core are lower than the safety criteria. Nuclear reactor analysis codes including MTR_PC package and MCNP5 were employed to carry out these calculations.     

Integral boundary layer heat transfer modeling for Tehran Research Reactor

In the present work the validity of applying the Boussinesq approximation in the analysis of natural convection heat transfer along nuclear fuel plates with large coolant channel aspect ratios is evaluated. The Boussinesq approximation is introduced into the integral boundary layer equations governing the system to describe the velocity and temperature distributions of the coolant in the cooling channels. The fuel plate temperature is related to the adjacent coolant fluid temperature by a fundamental law in conduction heat transfer. Air and water are considered as fluids. The coolant flow is assumed to be fully developed which is a convenient assumption for coolant channels having large aspect ratios. Obtained results indicate that the Boussinesq approximation is merely applicable over a limited range of coolant channel outlet fluid temperatures. The use of this approximation produces conservative estimation of the critical plate power for air flow and non-conservative estimation of the critical plate power for water flow.     

The effect of retarding torque during a flow transient for Tehran Research Reactor

The primary cooling system of the Tehran Research Reactor (TRR) has been analysed for a possible flow transient phenomenon caused by power cut-off. All the components of the TRR primary cooling loop that offer resistance to the coolant flow are physically modelled. Differential equations of motion for the coolant in the primary piping of the TRR and for the rotating parts of the centrifugal pump are then derived. The equation of flow motion is solved simultaneously with momentum conservation equation of the rotating parts of the pump which predicts the TRR pump speed during the flow transient. Electrical and mechanical losses are measured for the TRR three-phase induction motor in order to calculate the motor retarding torque during the event. The results of the present study are compared with the other similar primary loop results. The present model shows good agreement with the existing experimental and theoretical studies.     

Optimizing the performance of a neutron detector in the power monitoring channel of Tehran Research Reactor (TRR)

In this study, a general theoretical model is presented to calculate the current-voltage characteristics and associated sensitivity for a fission chamber. The chamber was used in a research nuclear reactor and a flux-mapping experiment was performed. The experimental current measurement in certain locations of the reactor was compared with the theoretical model results. The characteristic curves were obtained as a function of fission rate, chamber geometry, and chamber gas pressure. An important part of the calculation was related to the operation of the fission chamber in the ionization zone and the applied voltages affecting two phenomena, recombination and avalanche. In developing the theoretical model, we used the MCNP Monte Carlo code for fission rate and the SRIM program for ion-pairs computations. The theoretical model together with the above-mentioned codes was used to evaluate the effects of different applicable variations on the chamber's parameters.     

BNCT project at Tehran Research Reactor: Current and prospective plans

Boron Neutron Capture Therapy (BNCT) for brain tumor treatment is under development at the Tehran Research Reactor (TRR). This paper presents all current research activities that were performed during recent years as well as the prospective of BNCT research at TRR. The theoretical and experimental investigations show that TRR has a very good potential to consider it as a pilot facility for BNCT research in the Middle East and could be facilitated for clinical applications. In this way, there are some steps and also some challenges which are described in the paper.     

Uncertainty evaluation of calculated and measured kinetics parameters of Tehran Research Reactor

Effective delayed neutron fraction β_eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20 °C) and hot (fuel, clad and coolant temperature 65, 49 and 44 °C, respectively) states using MTR_PC computer code. The ratio of (β_eff)_i/(β_eff)_core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield (ν_d) and fission cross-section of ²³⁵U and ²³⁸U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β_eff and 5.7% for Λ in the cold state.     

Calculation,measurement and sensitivity analysis of kinetic parameters of Tehran Research Reactor

Effective delayed neutron fraction β_eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetic parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperatures equal to 20 °C) and hot (fuel, clad and coolant temperatures of 65, 49 and 44 °C, respectively) states using MTR_PC code. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. For the cold state, calculated β_eff and Λ by MTR_PC code are 0.008315 and 30.190 μs, respectively. In the hot state, these parameters being 0.008303 and 33.828 μs, respectively. The measured β_eff and Λ for the cold state (reactor power in the range of 100–200 W) being 0.008088 and 32.001 μs, respectively. The calculated and measured values are in good agreement. Relative percent errors are about 2.8% for β_eff and 5.7% for Λ which are smaller than the other reported results. In the third stage of the research, variations of β_eff and Λ vs. fuel enrichment are investigated in cold and hot states. Comparative analysis shows that both β_eff and Λ increase as fuel enrichment decreases. However, the variation rate of β_eff is not the same in the two conditions. β_eff in the hot state is larger than that calculated in the cold state when fuel enrichment is more than 83.91%, while the situation is vice versa for the enrichments smaller than the aforementioned value. Calculated neutron generation time shows normal behavior for all different fuel enrichments. All variables involved in kinetic parameters calculations (i.e., neutron fission cross section, fuel enrichment, etc.) are investigated theoretically to evaluate the results of calculations in cold and hot states. Variations of β_eff and Λ with fuel burnup are also studied.     

Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.     

Calculation of dpa rate in graphite box of Tehran Research Reactor(TRR)

Radiation damage is an important factor that must be considered while designing nuclear facilities and nuclear materials. In this study, radiation damage is investigated in graphite, which is used as a neutron reflector in the Tehran Research Reactor(TRR) core.Radiation damage is shown by displacement per atom(dpa) unit. A cross section of the material was created by using the SPECOMP code. The concentration of impurities present in the non-irradiated graphite was measured by using the ICP-AES method. In the present study the MCNPX code had identified the most sensitive location for radiation damage inside the reactor core. Subsequently, the radiation damage(spectral-averaged dpa values) in the aforementioned location was calculated by using the SPECTER, SRIM Monte Carlo codes, and Norgett,Robinson and Torrens(NRT) model. The results of ‘‘Ion Distribution and Quick Calculation of Damage'(QD)method groups had a minor difference with the results of the SPECTER code and NRT model. The maximum radiation damage rate calculated for the graphite present in the TRR core was 1.567 × 10~(-8) dpa/s. Finally, hydrogen retention was calculated as a function of the irradiation time.     

利用中国实验快堆生产放射性同位素的可行性研究

中国实验快堆（CEFR）不仅能进行各种燃料、材料辐照实验,也是放射性同位素生产的优良平台。本文对CEFR的辐照性能进行了描述,并利用计算程序对适宜在CEFR上生产的同位素³²P、³³P、³⁵S、⁸⁹Sr、¹⁴C、⁶⁰Co进行理论计算,得到了产量和比活度等参数。计算结果表明,在CEFR堆芯辐照可得到纯度极高的³²P、³³P、³⁵S,利用快中子的（n,p）反应可得到无载体的⁸⁹Sr,在CEFR反射层布置慢化材料可得到比活度较高的¹⁴C、⁶⁰Co。以上结果表明,在CEFR上生产同位素是可行的。     

反应堆一回路系统优化设计方案的可行性验证

采用优选运行参数和结构参数的方法,可达到降低核动力装置尺寸的目的。在优化设计方案投入制造前,有必要研究其在设计基准事故下的响应特性,以检验优化方案的可行性。采用RELAP5/MOD3.2程序研究现有一回路系统优化方案在完全失去厂外电、主给水丧失和小破口失水事故下的响应特性,并将安全设计准则参数与母型对比。结果表明：针对所研究的3种设计基准事故,优化方案各主要安全准则参数满足设计要求;优化方案可成功抵御这3类设计基准事故。     

Optimization of in-core fuel management strategy of Tehran Research Reactor (TRR) using MCNP-4C

In order to optimize fuel utilization in TRR, the method of fuel management is modified using MCNP-4C code system. An important parameter of fuel management is the uniformity of neutron flux distribution in the core region, which is obtained efficiently in the present strategy. This strategy is based on calculation of position factors and power densities utilizing MCNP simulations. This study shows that the core life time and average extracted burn up of spent fuel elements of TRR are improved significantly.     

脉冲反应堆动态参数研究

本文介绍了脉冲反应堆瞬变的动态特性的计算模型。计算结果与实验值的相对偏差约3%。     

铀氢化锆元件小型动力堆应用可行性分析

介绍了铀氢化锆燃料元件的主要性能和特点(尤其是热物理性能和堆内辐照性能),以及将铀氢化锆元件应用于动力堆所完成的一些研究概况.在此基础上,对铀氧化锆元件小型动力堆的技术可行性进行了论证和分析.结果表明:将铀氢化锆元件作为小型动力堆元件在燃料元件方面不存在严重的技术问题;使用细棒铀氢化锆元件的小型动力堆仍有较大的瞬发负温度系数,具有一定的固有安全性.     

研究堆周期保护装置堆上考验试验

周期保护装置是研究堆必备的设备之一,属于核安全一级设备.新研制的周期保护装置在安装到反应堆之前,必须在零功率堆或其它反应堆上进行堆上考验试验,以检验其周期保护和周期测量功能及测量精度.叙述了新研制的三套周期保护装置在堆上考验的试验设施、试验堆芯、试验内容和方法以及试验结果.试验结果表明:三套周期保护装置均满足技术指标的...     

Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown.Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.