Benchmark analysis of SPERT IV reactor

Benchmark calculations have been performed for SPERT IV D-12/25 core. Experimental data of the core was provided by International Atomic Energy Agency (IAEA). Combination of WIMS/D4 and CITATION codes has been used for performing the neutronic analysis of the reactor. Lattice calculations have been performed through WIMS/D4 while 3-dimensional reactor core has been modeled in CITATION. Ten energy groups were considered for these calculations. Energy wise microscopic cross-sections were generated for fuel, control absorber, control follower, guide tube, grid plate, reflector and structural regions separately of the core using WIMS/D4. Thermal neutron flux profiles at different axial and radial locations of the core were evaluated. Critical position of the control rods, excess reactivity, shut down margin, control rod worth, reactivity feed back coefficients and kinetic parameters of the core were estimated. Reasonable agreement has been found between experimentally determined and the calculated parameters.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρ_ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).     

Sensitivity analyses of the use of different reflector materials on the neutronics parameters of Tehran research reactor

In this paper, the effect of changes in neutron reflector type on neutronics parameters of Tehran research reactor is analyzed. In this study, at first, calculations for the HEU and LEU fuel configurations of the reactor core in which the light water is used as a neutron reflector in the core is done. Then, by using the reflectors such as graphite, beryllium and heavy water, changes on the neutronic parameters are examined. The results show that by altering the reflector, at HEU core configuration (compared with LEU), more changes appear in parameters such as neutron multiplication factor, average fast and thermal neutron flux, excess reactivity and shut down margin. Moreover, at LEU core configuration, changes are tangible specifically on parameters of cycle length and power peaking factor. In addition, the evaluated fuel temperature coefficient of reactivity is greater at HEU than LEU, while the temperature alteration of fuels represented greater influence on reactivity at LEU configuration.     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP

A neutronics feasibility study has been performed to determine the enrichment that would be required to convert a commercial Miniature Neutron Source Reactor (MNSR) from HEU (90.2%) to LEU (<20%) fuel. Two LEU cores with uranium oxide fuel pins of different dimensions were studied. The one has the same dimensions as the current HEU fuel while the other has the dimensions as the special MNSR, the In-Hospital Neutron Irradiator (INHI), which is a variant of the MNSR. The LEU cores that were studied are of identical core configuration as the current HEU core, except for potential changes in the design of the fuel pins. The following reactor core physics parameters were computed for the two LEU fuel options; clean cold core excess reactivity (ρ_ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, β_eff and prompt neutron lifetime, l_f). Results obtained are compared with current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of NIRR-1 in particular from HEU to LEU.     

Steady-state thermal hydraulic and safety analyses of a proposed mixed fuel (HEU&LEU) core for Pakistan research reactor-1

Pakistan Research Reactor (PARR-1) was converted from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, in 1992. The reactor is running successfully with an upgraded power level of 10 MW. In order to save money on the purchase of costly fresh LEU fuel elements, it is being thought to use some of the less burnt HEU spent fuel elements along with the present LEU fuel elements. In the present study steady-state thermal hydraulics of a proposed mixed fuel core (see Fig. 2) has been carried out. Results show that the proposed core, comprising of 24 LEU and 5 HEU standard fuel elements, with 4 LEU and one HEU control fuel elements, can be safely operated at a power level of 9.86 MW without compromising on safety. Standard computer codes and correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core and margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB).     

Homogeneous zones definition in deterministic codes and effect on computed neutronic parameters

The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividing it into homogenized zones, i.e. volumes within which the macroscopic nuclear properties are considered uniform. These codes have been extensively used and tested for several decades and are shown to perform well when they analyze reactor cores containing regions with relatively homogeneous distributions of fuel, moderator and absorbing materials. In this work, the sensitivity of computed key neutronic parameters to the partitioning of the reactor core in homogenized zones is examined. Application is made for a configuration of the Greek Research Reactor (GRR-1) core, which is pool type, fueled by slab-type fuel elements. For the calculations, the neutronic code system consisting of XSDRNPM (cell-calculations) and CITATION (core analysis) is used with two different definitions of homogeneous zones for the special/control fuel assemblies. The effect on computations of neutron flux distribution, void-induced reactivity and total control rod worth is examined based on corresponding measurements. It is shown that with a more appropriate partition in homogeneous zones, the agreement of computed results with measurements can be remarkably improved concerning mainly the neutron flux, while the control rods worth is the less affected quantity.     

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k_∞ for this methodology comparing the calculated and the experimental results.     

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.     

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (_lp$l_p$) and the effective delayed neutron fraction (_βeff${\beta }_{\mathit{eff}}$). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future.     

微型中子源堆低浓化堆芯的探讨

通过对２３５Ｕ富集度为１９．９％的ＵＯ２和Ｕ３Ｓｉ２－Ａｌ的弥散体２种燃料进行物理计算，从中筛选出了优化的堆芯方案，并对其静态物理参数，诸如有效倍增因子、绝对中子通量密度、上铍反射层反应性价值、反应性温度系数、控制棒价值等进行了计算。     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

高功率研究堆低浓化物理特性研究

应用ＦＧ２ＤＢ两维两群扩散燃耗程序和带６９群中子截面库的ＣＥＬＬ栅元少群参数程序，对高功率研究堆低浓化堆芯进行了物理计算。ＬＥＵ燃料元件的铀密度为３．６－７．２ｇ／ｃｍ３，包壳厚度为０．３８－０．５６ｍｍ。结果表明：改变燃料芯体铀密度或厚度在物理上相当；各堆芯方案的控制棒价值等运行安全有关参数都可以接受。部分计算结果被拟合成线性或二次关系式以便于应用。给出了各堆芯的最小临界值、剩余反应性、运行寿期、快热中子通量和积分通量等物理参数。分析这些参数后指出：当Ｕ－２３５含量提高２０％或更多时，ＬＥＵ堆芯与ＨＥＵ堆芯的主要物理性能相近，这时快中子通量几乎不受影响，热中子通量的下降率近似正比于元件Ｕ－２３５含量增加率。但由于ＬＥＵ堆芯运行寿期的延长，对一般同位素生产与燃料元件辐照考验不会有明显影响。     

Analysis of SHE Critical Experiments by Neutronic Design Codes for Experimental Very High Temperature Reactor

Analyses have been performed on various experiments conducted using the Semi-Homogeneous Experimental Assembly (SHE) to examine the accuracy of computer codes employed in the neutronic design of experimental Very High Temperature Reactor (VHTR). The neutronic design codes are DELIGHT-6 to obtain the neutron spectrum of a fuel cell and to produce group constants with burnup utilizing the nuclear data from ENDF-B/IV, CITDEGA to calculate the three-dimensional core performance considering the coupling effect between neutronic and thermohydraulic characteristics, and ANISN-JR and TWOTRAN-II for transport calculation. These codes are examined by the analysis on the integral quantities of effective multiplication factor, neutron flux distribution, burnable poison rod worth and control rod worth. The maximum degrees of disagreement with the relevant experiments are 0.57, 5, 7 and 5%, respectively.     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.     

Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

DF-VI快中子临界装置2号堆芯临界实验

DF-VI快中子临界装置在改造完成、堆芯发生了变化以后,进行了重新启动和一系列的实验测量。测量内容有：根据29次临界实验的数据对2号堆芯平均临界元件数和临界质量进行了计算;应用周期法和棒补偿法对控制棒价值进行了刻度;用逆动态反应性计对安全棒和安全块的价值进行了测量;对单根边缘元件价值和径向元件价值分布进行子测量。通过以上实验测量,确定了DF-VI快中子临界装置2号堆芯的主要安全运行参数。     

铝基碳化硼材料堆内辐照方案设计

铝基碳化硼是一种新型的乏燃料贮存格架用材料,为检验其辐照性能,需进行堆内辐照实验。本文从样品成分及形状、辐照罐结构、辐照位置等方面,对铝基碳化硼材料堆内辐照方案进行设计。经初步中子物理学和热工计算表明：在所选择的两个辐照孔道内进行辐照考验,试件所接受的累积γ射线照射剂量和相应的快中子积分注量均满足技术要求,且辐照罐样品入堆后对功率峰值因子、反应性、发热率等与堆运行安全相关因子的影响均在安全范围内。