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.     

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.     

The effect of high pressure test loop (HPTL) on the Reactivity of ET-RR-2 reactor

The second Egyptian Research Reactor ET-RR-2 is a multipurpose research reactor. It is an open pool type, with nominal power of 22 MW water-cooled. The reactor pool is designed to accommodate two fuel test loops mainly 500 and 20 KW loop in the reactor reflector to enable performing experiments on the behavior of fuel rods for nuclear reactors under their operating conditions. For that, inserted high-pressure test loop (HPTL) loaded with suggested CANDU type fuel element in the reactor core is important to achieve the above reason. From the neutronic safety point of view, it is necessary to study the mutual neutronic and reactivity effect between the reactor core and HPTL. This paper aimed at the study of the temperature coefficients of fuel and moderator of the CANDU type fuel element at different ²³⁵U enrichments, and the effect of HPTL on the reactor core reactivity. The effect of flooding the contact second shut down system (SSS) chamber with water and gadolinium nitrate on the reactor core reactivity in the presence of HPTL. All analysis was performed with the WIMSD4 and DIXY2 codes. This study shows that, an unacceptable change of reactor core reactivity was found due to the presence of the HPTL and the maximum inserted reactivity does not exceed 527 pcm at high possible ²³⁵U enrichment (10%).     

Burn-up effect on inherent safety parameters and reactivity insertion transient analysis of Pakistan Research Reactor-1

Burn-up dependent feedback coefficients of reactivity for the reference operating core of Pakistan Research Reactor-1 (PARR-1), have been calculated employing standard computers codes WIMSD/4 and CITATION. Fast reactivity insertion transient (1.5 $/0.5 s) is simulated at each burn step using computer code RELAP5/MOD3.4 and PARET. Calculation reveals that fuel temperature coefficient of reactivity is 1.77 %Δk/k/ΔT less negative while moderator temperature and void coefficients of reactivity are 7.74 %Δk/k/ΔT and 2.04 %Δk/k/ΔT more negative at end of cycle (EOC), respectively. Fast reactivity insertion transient analysis shows that due to larger value of prompt generation time (Λ), reactor response to transient is slow at EOC. Therefore peak power, maximum fuel centreline and clad temperature decrease as the fuel is burned. This is the sign of enhanced inherent safety with the burn-up of reference operating core of PARR-1. Removal of in-pile experiment accident has also been modelled in RELAP5/MOD3.4 and results in this study are compared with PARET.     

Revised calculation route for two group macroscopic cross-section library for RBMK-1500 type reactor

To increase the accuracy of predicted reactivity effects and coefficients for the unit equipped with a RBMK-1500 type reactor at Ignalina NPP, the calculation route used to generate the library of nuclear data constants applied in the neutronic/thermal hydraulic analysis has been updated with a modern version of the WIMS lattice code, WIMS8. The previously available two group library used with the QUABOX/CUBBOX-HYCA code, adapted to model the physical and nuclear processes in a RBMK-1500 reactor core, was created using the freely available WIMSD reactor physics cell code and its associated nuclear data library. In this article, the results that are obtained under the performance of the new two group cross-section library generated with WIMS8 for RBMK-1500 design core are presented. This discussion is mostly concentrated on the prediction of the key physics parameter for the RBMK type reactor, the void reactivity coefficient, as this parameter has been underestimated, especially at higher fuel irradiation.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

Comparative study of fuel burn-up and radioactive inventory for proliferation and proliferation resistant fuel lattices

A comparative study of fuel burn-up and radioactive inventory for the proliferation and proliferation resistant fuel lattices is carried out using the computer code WIMSD4. It is shown that by replacing the natural uranium metallic fuel and light water as coolant, such as used in the NRX/CIR, with the natural uranium oxide fuel and heavy water coolant, respectively, the core becomes proliferation resistant and environmental friendly as it produces about half the amount of ²³⁹Pu and presents about half the amount of activities associated with major actinidies as compared to the original NRX/CIR core. The infinite multiplication factors and the ²³⁹Pu produced in both fuel lattices are also compared.     

Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

International Atomic Energy Agency (IAEA) has recently released new WIMSD libraries based on current cross-section evaluations. Using these libraries the effect of different evaluated data sets on effective multiplication factor and neutron energy spectrum was studied with the help of 3D reactor simulation code CITATION. Simulation methodology adopted in this work was validated by analyzing IAEA 10 MW benchmark reactor.The k_eff values obtained using all newly released libraries are within 0.45% to the experimental value, while the old library released in 1981 resulted in calculated value 1.05% larger than experimental. The flux spectrum obtained for standard fuel element using 3D modeling is smaller in fast energy range and higher in thermal energy range than is calculated using the 1D model for the standard cell. In the flux trap, differences of about −4% to 13% were found in thermal flux using the newly released libraries as compared to that obtained using 1981 WIMSD library. The major differences in the flux spectra between newly available libraries and the 1981 WIMSD library in thermal energy range are due to the differences in cross-sections of hydrogen bound-in-water. The use of only newly available cross-sections of hydrogen bound-in-water with 1981 WIMSD library resulted in significant improvement in value of k_eff as well as in the flux spectrum. Moreover the differences among new libraries in the thermal energy range are also due to these cross-sections. Difference in fission spectra from different libraries is responsible for differences of flux spectra in the fast energy range. These differences in flux are reduced significantly in the fast energy range by only replacement of fission spectra.     

Fission MOLY production at PARR-1 using LEU plate type target

Neutronic analyses and depletion calculations have been performed for the production of ⁹⁹Mo at PARR-1. Analysis has been performed with 20% enriched ²³⁵U target bearing plate type aluminized fuel (U-Al). Target (target holder and fuel plates) design contains three fuel plates and two aluminum dummy plates. Neutronic calculations were carried out for core at the beginning of equilibrium cycle of Pakistan Research Reactor-1 (PARR-1). Target analysis was performed by irradiating it at a location of maximum thermal flux available in the core. For this purpose, irradiation was performed at five different axial planes of the central water box facility. Thermal neutron flux profiles were also studied at different axial positions in available irradiation locations. Computer code WIMSD/4, a transport theory lattice code was employed for the generation of 10 group microscopic cross-sections. Diffusion theory code CITATION was utilized for three-dimensional modeling of the core. It was observed that from reactivity point of view, insertion or removal of target from the core will not affect the safety of reactor. Maximum heat flux in the target would be 102.68 W/cm² which is below the point of onset of nucleate boiling. However, forced flow is required to avoid initiation of nucleate boiling. The computer code ORIGEN2 was employed for depletion calculations. Analysis was performed for 100 Ci activity of ⁹⁹Mo. After 100 days decay, waste activity will be less than 1 Ci and it will not pose any problem for handling radioactive waste.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Design concept of a small nuclear reactor for large-diameter NTD-Si using a conventional PWR full-length fuel assembly

A design concept for a small nuclear reactor dedicated to large-diameter neutron transmutation doping silicon (NTD-Si) is proposed. Conventional PWR (Pressurized Water Reactor) full-length fuel assembly is used to assure stable and reliable supply of fuel. Criticality, neutron transportation, and core burn-up calculations are performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the proposed reactor can be critical over 18 years, and excess reactivity can be suppressed by a combination of Gd₂O₃ burnable poison and soluble boron. Preliminary steady-state single-channel thermal hydraulic analysis showed that heat removal from core is possible under 1 atm operating pressure. Si ingots up to 30 cm in diameter can be irradiated in the reactor irradiation channels, and the uniform irradiation condition can be achieved for a large-diameter Si ingot.     

Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% ²³⁵U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% ²³⁵U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

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.     

COSINE软件包组件参数计算程序LATC共振模块研制与验证

共振计算在反应堆物理计算中具有非常重要的意义。本文基于压水堆组件的特点,开发了用于LATC组件计算程序的共振模块。该共振模块采用成熟的等价理论模型,首次碰撞概率采用二项有理近似,可读取WIMSD格式和WIMSD改进型格式的多群截面数据库,同时可直接提供用于LATC输运计算的宏观截面数据。针对程序运行过程中涉及的大量截面数据计算与传递,对数据存储结构进行了优化,使计算速度有较大提高。基于LATC组件程序对该模块进行了初步验证分析,并与组件程序DRAGON进行了比较,初步数值结果表明共振模块有良好的计算精度,能满足当前轻水堆组件设计的要求。     

VVER-1000 neutronics calculation with ENDF/B-VII data

The initial release of the ENDF/B-VII nuclear data library is verified for VVER-1000 reactors. For neutronics calculation, the MCNP code based on the Monte-Carlo method is applied. Continuous-energy cross-sections for use with MCNP are calculated with the NJOY code. Isotopics for burned fuel is calculated with the WIMSD code. Calculated criticality, pin-to-pin power distribution, time-dependent critical concentration of soluble boron, worth of the control rods, average fuel assembly powers and time-dependent axial power distribution are compared to the corresponding experimental values for both zero-power VVER-1000 model, created at the LR-0 experimental facility, and the first fuel cycle of a real VVER-1000 reactor. For all of these parameters, neutronics calculation with ENDF/B-VII is in good agreement with the measurement. Moreover, for VVER-1000 neutronics calculation, ENDF/B-VII provides better results than ENDF/B-VI.     

Computational investigations of the dynamical characteristics of the core of a periodic pulse reactor in a system with cascade neutron multiplication

The core dynamics of a fast reactor in a cascade reactor system operating in a periodic-pulse regime are examined. A model of a BN-600 fuel element is used as a computational model. Computational studies of the neutron kinetics processes in a fast rector-subcritical assembly system and the thermal dynamics of a fuel element in the core of a periodic-pulse reactor are performed. Estimates are made of the service life of a fuel element operating in a regime with repeating pulses and a number of heat loads that is admissible from the standpoint of the fatigue strength of the element.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 260–269, October, 2004.     

压水堆平衡堆芯钍铀燃料循环初步研究

建立WIMSD5-SN2-CYCLE3D和CASMO3-CYCLE3D物理分析系统作为钍铀燃料循环研究工具.以大亚湾第1机组压水堆为参考堆型,不改变反应堆栅元、组件和堆芯的结构与几何尺寸,设计出含36根钍棒、4.2#5U富集度的新型含钍组件,并对含钍组件和3.2%富集度的铀组件进行中子学计算和分析.模拟并分析了大亚湾压水堆12个月换料从初始循环到铀钚平衡循环的换料过程.再从平衡铀堆芯出发,逐步加入含钍组件代替铀组件,对铀钚平衡循环到钍铀平衡循环的换料过程进行了模拟与分析.计算结果表明:钍铀平衡循环比铀钚平衡循环每天节省裂变核素质量约18.4%,并减少了长寿命放射性核废料的产生.不利因素是使得循环长度减少90EFPD,缩短了换料周期,增加运行费用,并给燃料管理、安全控制以及乏燃料的处理带来困难.建议提高组件的235U富集度,在压水堆上进行钍利用研究.     

中国实验快堆热工参数的自适应BP神经网络预测方法研究

中国实验快堆（CEFR）堆芯的热工参数是否超出限值是评价反应堆安全运行的标准。本文针对燃料包壳最高温度预测问题,通过堆芯子通道分析程序COBRA生成数据样本后,开发基于BP神经网络自适应算法的智能预测程序,对于特定的单盒组件,仅需给出堆芯进口功率和流量,即可实现燃料包壳最高温度的快速准确预测。结果表明,与COBRA相比,在大规模重复性计算的场景下,自开发程序能节约大量计算时间和算力,提高燃料包壳设计和CEFR运行时的操作效率。实验分析得出BP神经网络方法的最大相对误差不超过6%,平均预测相对误差不超过3%,计算效率提升至原程序的300倍,网络模型的预测精度高,且易推广至实验快堆其他参数预测,具有很大的应用前景。     

Four energy group neutron flux distributions in the Syrian Miniature Neutron Source Reactor using the WIMSD4 and CITATION codes

A 3-D (R, θ, Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the pointwise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation sites with relative differences less than 7% and 5%, respectively.