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。     

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.     

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.     

Research on MNSR LEU Convertion

正China Institute of Atomic Energy(CIAE)has designed and constructed prototype Miniature Neutron Source Re a c tor(MNSR)in 1984.Then CIAE developed it,and constructed commercial MNSR.After that,3 commercial MNSRs were built in China and 5commercial MNSRs were built abroad.These MNSRs     

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.     

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).     

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.     

Design and experience of HEU and LEU fuel for WWR-M reactors

Thin-walled WWR-M5 fuel elements were designed and manufactured and have been used successfully for 16 years; they contain twice as much uranium-235 as the WWR-M2 and WWR-M3 fuel elements. The fuel elements have been optimized with regard to their neutron physics and thermal-hydraulic parameters and fuel consumption has been minimized. The mean specific power in the core of the WWR-M reactor was raised to 230 kW l⁻¹, the measured maximum volume thermal specific power was 900±100 kW l⁻¹ and the surface specific power was 136±15 W cm⁻². The WWR-M5 fuel elements enable the power of the WWR-M pooltype reactor to be raised to 30 MW while simultaneously increasing the number of cells in the core available for experimentation by a factor of approximately two and reducing fuel element consumption. Reactor tests of WWR-M fuel elements with reduced fuel enrichment (36 and 21%) were carried out for a meat uranium density up to 2–3 g cm⁻³. Conversion of WWR-SM-type reactors to these fuel elements did not lead to a loss in reactivity and enabled their power to be increased to 20–30 MW.     

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     

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.     

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.     

Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl₄-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U²³⁵. Standard computer code PARET/ANL (version 1992) (Obenchain, 1969) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface & maximum fuel centerline temperatures; and peak power & corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR) (Qazi et al., 1994). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% & 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.     

Preliminary Design of Neutron Dose Measurement Device for Cosmic-Ray

In order to be applied in cosmic ray, a device was designed which uses a spherical moderator and two types of proportional counters. One is the spherical counter which is imbedded at the center of the sphere     

Preliminary Design of Neutron Dose Measurement Device for Cosmic-Ray

In order to be applied in cosmic ray, a device was designed which uses a spherical moderator and two types of proportional counters. One is the spherical counter which is imbedded at the center of the sphere moderator. This counter is called as an inner detector. The other six counters is the tube counter. Each is located close to the moderator surface and these counters are called as an outer detector.     

微堆运行性能的提高

介绍了深圳大学微堆几年来在安全运行的基础上为提高微堆的运行性能所进行的技术改进。目前,微堆在额定功率下的最大可运行时间从约８ｈ提高到约４０ｈ,运行性能大为提高,为中短寿命放射性同位素制备和化分析应用展现了新的前景。     

Preliminary Design of Source Neutrons for Detecting Explosives by Thermal Neutron

<正>The explosive detection method of thermal neutron activation(TNA)uses the thermal neutron interacting explosive of rich elements(H,N),after(n,γ)reaction,emitγrays with characteristic energy(2.223 MeV,10.835 MeV),through spectrum analysis the content of reaction element can be determined,detecting the presence of explosives.Based on the D-D neutron     

Preliminary Design of Secondary Neutron Source for China Experimental Fast Reactor

正To improve economic efficiency during the physical start-up of China Experimental Fast Reactor(CEFR),we choose Sb-Be neutron source as the design object,the feasibility of using sec-     

Post-shutdown decay power and radionuclide inventories in the discharged fuels of HEU and potential LEU miniature neutron source reactors

Assessment of fission product and actinide content along with the time variation of decay power of discharged fuels of both HEU and LEU cores of MNSRs have been carried out for once-through cycle using the ORIGEN2 computer code. The results for the LEU core have been compared with the corresponding values for the current HEU core of MNSRs. For the HEU and the potential LEU UO₂, U-9Mo discharged fuels, the ORIGEN2 computed isotopic and total activity values have been found in good agreement with the corresponding results obtained by using the WIMSD4 code. All three MNSR fuels show fission product dominated activity behavior for post-shutdown periods up to about 10³ years during which, the total activity decreases by as much as 10⁶ times. The residual actinide activity shows smaller variations as the three discharged fuels decay thru 10⁶ years. The time variation of the decay power follows the same behavior as the corresponding total activity values during the fission product dominated period. A decrease from initial values of 154.76, 162.6,160.39 W to the final values 9.35 × 10⁻⁵, 2.1 × 10⁻³, 1.7 × 10⁻³ W has been found for the standard HEU, and potential UO₂, U-9Mo LEU fuels correspondingly during this time. The standard HEU fuel shows smallest decay power values while the UO₂ and U-9Mo LEU fuels have comparable values for time spans from 10³ to about 10⁶ years.     

Core Thermohydraulic Design with 20% LEU Fuel for Upgraded Research Reactor JRR-3

This paper presents the outline of the core thermohydraulic design and analysis of the research reactor JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% low enriched uranium (LEU) plate-type fuel. For the condition of normal operation, the upgraded JRR-3 core is planned to be cooled by two cooling modes of forced-convection at high power and natural-convection at low power. The major feature of core thermohydraulics is that at the forced-convection cooling mode the core flow is a downflow, under which fuel plates are exposed to a severer condition than an upflow in cases of operational transients and accidents. The core thermohydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margins both against the onset of nucleate boiling (ONR) not to allow the nucleate boiling anywhere in the core and against the departure from nucleate boiling (DNB). The safety margins against ONB and DNB were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the ONB, and the minimum DNB ratio (ratio of DNB heat flux to the maximum heat flux) was evaluated to be about 2.1, which gives a sufficient margin against the DNB. The core thermohydraulic characteristics were also clarified for the natural-convection cooling mode.