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).     

Preliminary Study of MNSR With LEU Fuel

The Miniature Neutron Source Reactor （MNSR） is a low power research reactor. It was developed, designed and manufactured by China Institute of Atomic Energy （CIAE）, with high enrichment of ^235U 90% UAl4 alloy fuel. The first Prototype MNSR reached full power in 1984. Till now, three domestic commercial MNSRs have been built in Shenzhen, Shandong and Shanghai, another five commercial MNSRs in Pakistan, Iran, Ghana, Syria and Nigeria, last three were recommended by IAEA.     

CIAE Succeeds in LEU Modification of Ghana's MNSR

<正>On August 29~(th),Ghana MNSR’s High Enriched Uranium(HEU)fuel has transported back from Ghana to China safely and smoothly.So far,the Ghana MNSR LEU conversion project led by CIAE was successfully completed.The successful implementation of Ghana MNSR,the first one which has done LEU conversion abroad,is an important     

Determination of enrichment of recycle uranium fuels for different burnup values

Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000–50,000 MWd/tonU.     

A comparative neutronic study of the standard HEU core and various potential LEU alternatives for a typical MNSR system

A comparative study has been performed for neutronic analysis of highly enriched in uranium (HEU) and potential low enriched in uranium (LEU) cores for the Pakistan Research Reactor-2 (PARR-2) taken as a typical miniature neutron source reactor (MNSR) system. The group constant generation has been carried out using transport theory code WIMS-D4 and a detailed five-group RZ-model has been used in the CITATION code for multigroup diffusion theory analysis. The neutronic analysis of the 90% HEU reference and potential LEU alternative: UO₂, U₃Si₂ and U9Mo, cores has been carried out yielding 11%, 20.7% and 14.25% enrichments with corresponding values of excess reactivity: 4.33, 4.30 and 4.07 mk. These results have been found in good agreement with recently reported Monte Carlo-based transport theory calculations. The diffusion theory-based calculated values of thermal flux profiles for axial as well as for radial directions have been found to agree well with the corresponding experimental measurements. The UO₂-based LEU core has been found having flux spectrum closest to the reference core while U9Mo core has significantly harder flux spectrum at irradiation site.     

The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU

The Nigerian Research Reactor-1 (NIRR-1) falls in the category of Miniature Neutron Source Reactors (MNSR) using a fuel of 90% HEU. It is therefore desirable to convert it from this enrichment to LEU (less than 20%) in conformity with the new global trend of making research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for non-peaceful purposes. In this work, we have developed a computational scheme based on WIMS and CITATION that would theoretically achieve this objective as easily as possible. The scheme systematically reduces the enrichment from 90% (or any other initial values) to less than 20% in steps of 5% or any desired percentage variation. Two fuel types (UAl₄ and UO₂) are considered in here, while maintaining the size and geometry of the core as well as the excess reactivity (between 3.5 and 4 mk). Our results show that the U-235 loading increases sharply as enrichment decreases. It has also been noticed that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops sharply to 998 g for UAl₄ fuel and 946 g for UO₂ fuel. Below the enrichment of 5%, the operation of NIRR-1 with both fuel types can be considered unrealistic as this requires structural adjustment which the work tries to maintain constant.     

Reactivity temperature coefficients for the HEU and LEU fuel of the MNSR reactor

Analysis of the Reactivity Temperature Coefficients of the Miniature Neutron Source Reactor (MNSR) for normal and accidental conditions (above 45 °C) using HEU-UAl₄ and the LEU: U₃Si, U₃Si₂ and U9Mo fuel were carried out in this paper. The Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and Moderator Density Coefficient (MDC) were calculated using the GETERA code. The contribution of each isotope presented in the fuel cell was calculated for the temperature range of 20 °C–100 °C at the beginning of the core life. The average values of the FTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were found to be: −2.23E-03, −1.85E-02, −1.96E-02, −1.85E-02 mk/°C respectively. The average values of the MTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −8.91E-03, −1.24E-04, −4.70E-03, 2.10E-03 mk/°C respectively. Finally, the average values of the MDC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −2.06E-01, −2.03E-01, −2.04E-01, −2.03E-01 mk/°C respectively. It's found also that the dominant reactivity coefficient for all types of fuel is the MDC.     

高额超热中子束流治疗孔道低浓化BNCT堆初步设计方案

根据硼中子俘获治疗(BNCT)中子源的要求,在高浓铀为燃料的微型反应堆(MNSR)的基础上,以富集度19.5%的UO2为燃料,将其堆芯低浓化并且添加水平超热中子束流治疗孔道,开展超热中子束流BNCT堆堆芯低浓化初步设计。计算BNCT堆的超热中子注量率、单位超热中子注量的快中子剂量率、单位超热中子注量的γ光子剂量率、超热中子注量与热中子的注量之比、中子束流能谱等关键参数。结果表明,该设计可以得到优良的超热中子束流。     

A new method for uranium enrichment

A new method for uranium enrichment is presented. This method is based on exciting ²³⁸U to its isomeric levels at 2557.6 and 2557.6 + x keV. These isomeric levels have SF decay modes. Enrichment is thus obtained by fissioning only the ²³⁸U isotopes. Three ways for exciting ²³⁸U to ^238mU are discussed.     

Preliminary Research for Conversion of HEU to LEU for MNSR Design of In-Hospital Neutron Irridiator

The In-Hospital Neutron Irradiator （IHNI） with reactor power of 30 kW is specifically designed for Boron Neutron Capture Therapy （BNCT）. The reactor is an under moderated reactor of pool-tank type, and UO2 with enrichment of 12.5% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection.     

Design calculation of an epithermal neutronic beam for BNCT at the Syrian MNSR using the MCNP4C code

This article describes the design calculation of an epithermal neutronic beam for the boron neutron capture therapy at the Syrian MNSR by using the MCNP4C code and ENDF/B-V cross-section library. To produce a high flux of epithermal neutrons at the beam exit, the moderator/filter from Al, Cd, Fluental and Bi was used with Pb as reflector for neutrons along the beam. In addition, the Bi lined collimator with Li₂CO₃-PE and Pb at the end. The calculated beam parameters under 30.0 kW of reactor power at the beam exit are Ф_epi = 2.83 × 10⁸ n/cm² s, D_f/Ф_epi = 7.98 × 10⁻¹¹ cGy cm²/n, D_γ/Ф_epi = 1.70 × 10⁻¹¹ cGy cm²/n, Φ_epi/Φ_the = 0.05 and J_n⁺/Ф_n = 0.77. As well as, the calculated values of the advantage depth and advantage ratio are 7.51 cm and 3.49, respectively. If such beam was built into the Syrian MNSR the scientific applications of the reactor would increase.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

A numerical neutronics benchmark study for inert matrix plutonium fuels in uranium dioxide and mixed plutonium-uranium dioxide environments

A numerical benchmark exercise has been under way for comparing the results of different calculational methods/data sets used for the analysis of light water reactor (LWR) configurations employing Pu inert matrix fuels (IMFs). The first phase of the exercise was devoted to infinite arrays of identical IMF cells. The main feature investigated in the second phase has been the influence of the neutron spectra in UO₂ and MOX cores on individual IMF cells. Phase 3 is concerned with the more realistic situation of an IMF assembly surrounded by UO₂ assemblies. Significant discrepancies have been observed for power peaking effects and delayed neutron parameters in Phase 2. In Phase 3, neutron balance differences for the IMF, particularly at EOL, are found to be significantly larger than were observed in Phase 1.     

铀离心浓缩厂铀丰度在线监测仪本底自动测量方法研究

铀丰度在线监测仪是对铀浓缩厂工艺管道中UF6气体235U丰度进行在线监测的装置,本底是其核心关键技术指标,直接关系到丰度值的测量精度。原有本底测量方法需监测仪停止工作,人工将容器内的气体抽空进行测量。而本底自动测量方法通过改变测量容器内UF6气体的压力,用Na I(TI)探测器测量容器内UF6气体中235U发射的特征γ射线,利用压力传感器测量容器内UF6气体的压力值,最后对不同压力下的数据进行拟合获得监测仪的本底。实验结果表明,采用本底自动测量方法,监测仪铀丰度在线监测结果的相对标准偏差小于0.30%,与气体质谱计测量结果的最大相对偏差小于0.25%,表明该方法测量本底的准确度高;监测仪本底测量由软件自动完成,提高了监测仪的自动化程度,增强了监测仪的适用性。     

A feasibility study for a heavy-ion hybrid fusion-fission reactor using low-gain targets

A numerical study has been made which demonstrates the physics and engineering feasibility of a mixed fission-fusion inertial confinement power reactor. The system is a modular concept based on a reactor chamber using low-gain single-shell D-T pellets directly driven by a heavy-ion accelerator. The blanket is natural UC fuelled, with Li₂O as a T breeder and is cooled with pressurized He gas. Each chamber has a neutron first-wall loading of 0.1 MW m⁻² and a power output of 175 MW(th). As a 10-chamber system driven by a 25 Hz heavy-ion accelerator, the reactor would have a total output of 1.75 GW(th) with a structural materials lifetime of 15–20 yr.     

Simulation of delayed neutrons using MCNP

Accurate modeling of the delayed neutron response in a fission process has been a desired capability for MCNP^TM (Briesmeister, 2000). After a year of data library and code development, a delayed neutron feature has now been incorporated into the latest version of MCNP, 4C. In this work, a validation of the integrated delayed neutron model is performed by comparisons to an analytic solution and experimental results.     

A conceptual design study of proliferation-resistant PWR fuel using reprocessed uranium

A conceptual design study was carried out to enhance proliferation-resistant nature of current light water reactor fuels. Main features of the proliferation-resistant fuel design are adoption of alloy instead of oxide and utilization of enriched reprocessed uranium (10 wt% ²³⁵U). Major dimensions of the fuel assembly were not changed because of thermal-hydraulic considerations and back-fittability to current PWRs. Its smaller ²³⁸U inventory reduces generation of plutonium and ²³⁶U in the reprocessed uranium promotes generation of ²³⁸Pu that has large decay heat. The assembly calculation results of the fuel indicated that the fuel has good proliferation-resistant nature in the viewpoint of decreased plutonium generation, worse plutonium composition and increased decay heat. Neutronic analyses of an equilibrium core loaded with the proliferation-resistant fuels were carried out and calculation results indicate that variations of major core safety parameters are not very large. Therefore, basic feasibility of the proliferation-resistant fuel design using reprocessed uranium was confirmed in the course of this study.     

Dose calculations using artificial neural networks: A feasibility study for photon beams

Direct dose calculations are a crucial requirement for Treatment Planning Systems. Some methods, such as Monte Carlo, explicitly model particle transport, others depend upon tabulated data or analytic formulae. However, their computation time is too lengthy for clinical use, or accuracy is insufficient, especially for recent techniques such as Intensity-Modulated Radiotherapy. Based on artificial neural networks (ANNs), a new solution is proposed and this work extends the properties of such an algorithm and is called NeuRad.Prior to any calculations, a first phase known as the learning process is necessary. Monte Carlo dose distributions in homogeneous media are used, and the ANN is then acquired. According to the training base, it can be used as a dose engine for either heterogeneous media or for an unknown material. In this report, two networks were created in order to compute dose distribution within a homogeneous phantom made of an unknown material and within an inhomogeneous phantom made of water and TA6V4 (titanium alloy corresponding to hip prosthesis).All NeuRad results were compared to Monte Carlo distributions. The latter required about 7 h on a dedicated cluster (10 nodes). NeuRad learning requires between 8 and 18 h (depending upon the size of the training base) on a single low-end computer. However, the results of dose computation with the ANN are available in less than 2 s, again using a low-end computer, for a 150×1×150 voxels phantom. In the case of homogeneous medium, the mean deviation in the high dose region was less than 1.7%. With a TA6V4 hip prosthesis bathed in water, the mean deviation in the high dose region was less than 4.1%.Further improvements in NeuRad will have to include full 3D calculations, inhomogeneity management and input definitions.     

ECRH for JET: A feasibility study

For JET to fulfil its mission in preparing ITER operation, the installation of an electron cyclotron resonance heating system on JET would be desirable. The study described in this paper has investigated the feasibility of installing such a system on JET. The principal goals of such a system are: current drive over a range of radii for NTM stabilization, sawtooth control and current profile tailoring and central electron heating to equilibrate electron and ion temperatures in high performance discharges. The study concluded that a 12 gyrotron, 10 MW, system at the ITER frequency (170 GHz) adapted for fields of 2.7–3.3 T would be appropriate for the operation planned in JET. An antenna allowing toroidal and poloidal steering over a wide range is being designed, using the ITER upper launcher steering mechanism. The use of ITER diamond windows and transmission line technology is suggested while power supply solutions partially reusing existing JET power supplies are proposed. Detailed planning shows that such a system can be operational in about 5 years from the time that the decision to proceed is taken. The cost and required manpower associated with implementing such a system on JET has also been estimated.