Physical particularities of nuclear reactors using heavy moderators of neutrons

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using ²³³U as a fissile nuclide and ²³²Th and ²³¹Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

基于~(252)Cf源驱动噪声分析法的高浓缩铀部件外加反射层效应

针对高浓缩铀部件外加的反射层材质及厚度之效应可以巧妙避开核查这一问题,依据~(252)Cf中子源驱动噪声分析法原理,采用蒙特卡罗方法模拟研究了高浓缩球形金属铀部件的不同材料、厚度的反射层的效应,获得了相应的时间关联符合计数分布和中子产额。研究结果表明,对同一材料的反射层,反射层厚度愈大,中子产额愈大,即反射效果愈好。对于同一厚度的反射层,反射层材料的密度愈大,中子产额愈大,反射敷果愈好。     

Effective conditions for the neutron flux density at axial boundaries of the core

Analytical expressions for elements of the triangular matrix of effective conditions at the boundary of the core with a multiregion reflector are derived in the few-group diffusion approximation. The developed technique is verified using the example of fuel assemblies of a light-water reactor with an intermediate neutron spectrum.     

基于252Cf源驱动噪声分析法的高浓缩铀部件外加反射层效应

针对高浓缩铀部件外加的反射层材质及厚度之效应可以巧妙避开核查这一问题,依据252Cf中子源驱动噪声分析法原理,采用蒙特卡罗方法模拟研究了高浓缩球形金属铀部件的不同材料、厚度的反射层的效应,获得了相应的时间关联符合计数分布和中子产额。研究结果表明,对同一材料的反射层,反射层厚度愈大,中子产额愈大,即反射效果愈好。对于同一厚度的反射层,反射层材料的密度愈大,中子产额愈大,反射效果愈好。     

A conceptual high flux reactor design with scope for use in ADS applications

A 100 MWt reactor design has been conceived to support flux level of the order of 10¹⁵ n/cm²/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D₂O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 10¹⁴ n/cm²/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.      

On the Use of Depleted Uranium Dioxide as a Neutron Filter

A simple additive formula is given which, for neutron energies in the range 10^-4 < E<10 eV, permits calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections of a crystalline material as a function of its crystal constants and temperature. Computer codes PUO and SUO were developed to calculate the total attenuation of reactor neutrons through poly- and mono-crystalline samples, respectively. The codes were applied to calculate the total cross-section of polycrystalline uranium and thorium dioxides in the energy range from 4 meV to 1 eV. The obtained agreement between the calculated values and available experimental values justifies the applicability of the used formula and the computer codes. A feasibility study on using depleted polycrystalline uranium dioxide as a cold neutron filter and the single crystal as a thermal neutron filter is given. The optimum crystal parameters and thickness for efficiently transmitting the thermal reactor neutrons while strongly attenuating both fast neutrons and -rays accompanying the thermal ones is also given.     

Fission decay properties of ultra neutron-rich uranium isotopes

The fission decay of highly neutron-rich uranium isotopes is investigated which shows interesting new features in the barrier properties and neutron emission characteristics in the fission process. ²³³U and ²³⁵U are the nuclei in the actinide region in the beta stability valley which are thermally fissile and have been mainly used in reactors for power generation. The possibility of occurrence of thermally fissile members in the chain of neutron-rich uranium isotopes is examined here. The neutron number N = 162 or 164 has been predicted to be magic in numerous theoretical studies carried out over the years. The series of uranium isotopes around it with N = 154–172 are identified to be thermally fissile on the basis of the fission barrier and neutron separation energy systematics; a manifestation of the close shell nature of N = 162 (or 164). We consider here the thermal neutron fission of a typical representative ²⁴⁹U nucleus in the highly neutron-rich region. Semiempirical study of fission barrier height and width shows that ²⁵⁰U nucleus is stable against spontaneous fission due to increase in barrier width arising out of excess neutrons. On the basis of the calculation of the probability of fragment mass yields and the microscopic study in relativistic mean field theory, this nucleus is shown to undergo exotic decay mode of thermal neutron fission (multi-fragmentation fission) whereby a number of prompt scission neutrons are expected to be simultaneously released along with the two heavy fission fragments. Such properties will have important implications in stellar evolution involving r-process nucleosynthesis.      

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Using a one-dimensional (1D) neutronics model, the neutronics performance in the China fusion engineering test reactor (CFETR) with latest design dimensions of vacuum vessel is calculated under the 2GW fusion power. The shielding effect of neutron reflecting material ZrH₂ on neutrons is calculated, and it is found that the 20cm reflector can shield 94.3% neutron fluence and 94.9% neutron nuclear heat. Meanwhile, the minimum shield blanket thickness corresponding to different neutron wall loads is calculated when CFETR is operated at 10FPY (full power year) and 20FPY. The results show that the minimum shield blanket thickness are 44cm, 53cm, and 65cm corresponding to the neutron wall loads with 1.0MW·m⁻², 1.5MW·m⁻², and 2.5MW·m⁻² respectively after the device is operated at 10 FPY; whereas the shielding blanket needs to be thicker in the radial direction to meet the neutron shielding requirements after the device is operated at 20FPY. The optimized size of the shielding blanket provides a significant reference for the design of CFETR advanced blanket.     

2GW功率下CFETR中子屏蔽的初步中子学研究

在中国聚变工程实验堆(CFETR)真空室最新设计尺寸下,利用蒙特卡洛中子输运程序(MCNP)建立一维中子学模型,在2GW 的聚变功率下进行了计算。分析了中子反射材料ZrH₂ 对中子的屏蔽效果,发现200mm 的反射层可以屏蔽94.3%的中子通量和94.9%的中子核热。研究CFETR 在运行10 个满功率年(FPY)和20FPY 后,对应不同中子壁载荷的最小屏蔽包层厚度。结果显示,装置运行10FPY 后中子壁载荷在1.0MW·m⁻²、1.5MW·m⁻²、 2.5MW·m⁻² 时所对应的最小屏蔽包层厚度分别为44cm、53cm、65cm;而在装置运行20FPY 后,则需要在径向方向更厚的屏蔽包层才能满足中子屏蔽要求。屏蔽包层的尺寸优化将为目前阶段的CFETR 先进包层设计提供参考。     

Project of the ultracold and cold neutron source at the WWR-M reactor with superfluid helium as a moderator

The WWR-M reactor of the Petersburg Nuclear Physics Institute provides a unique opportunity for creating conditions of low radiative heat release (～4 × 10^?3 W/g) at a sufficiently high neutron flux (～3 × 10¹² neutrons/(cm² s)). This opportunity can be implemented in the reactor thermal column, which represents a 1-m-diameter channel adjacent to the reactor core. This diameter of the channel allows the arrangement of the core gamma shielding made of bismuth (15 cm thick), a graphite premoderator (300 dm³) at a temperature of 20 K, and a converter with superfluid helium (35 dm³) at a temperature of 1.2 K. Calculations show that the heat release in the source (20 W) can be removed by pumping helium vapor, and the density of ultracold neutrons in an experimental trap will be ～10⁴ neutrons/cm³, which is higher than that of existing sources of ultracold neutrons by two to three orders of magnitude.     

Franco-Russian comparison of mixed neutron and gamma radiation field dosimeters at the Silène reactor

This paper gives the results of dosimetry measurements carried out in the Silène reactor at Valduc (France) with neutron and photon dosimeters in mixed neutron and gamma radiation fields, in the frame of a Franco-Russian comparison of dosimeters. Neutron dosimetry was supplied by passive semiconductors, activation detectors and nuclear track detectors. For photon dosimetry, thermoluminescent and passive semiconductor detectors were used. The experiments were located at 3 m from the reactor core, in free air and also at the front and back of a tissue-equivalent phantom. The pulse operating mode of the reactor was used to simulate a criticality accident with solid fissile material, while the free evolution mode simulated a criticality accident in a fissile solution. The photon absorbed dose showed a slight increase on entering the phantom compared to measurements in free air, probably due to backscattering by the phantom. At the rear of the phantom, the neutron kerma was four times lower than on the front, whereas the photon dose was only two times lower. The heterogeneity of dose inside the phantom was far greater for neutrons than for photons.     

Multiple recycling of fuel in prototype fast breeder reactor

In a thermal neutron reactor, multiple recycle of U-Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.      

Optimization Study of the IBR-2 Reactor

The article considers the neutronics aspect of the IBR-2 reactor optimization: whether it is possible in theory to create an IBR-2-type reactor with a neutron flux in beams above the existing 0.5 × 1013 n/(cm2 s). The calculations have shown that the thermal neutron flux theoretically can be increased to (2.0−2.5) × 1013 n/(cm2 s), but only with a complete change in the reactor design: reducing the core volume, replacing the fuel type with a denser one, and changing the beam extraction system from radial to tangential. The technical implementation of these requirements is currently a challenge.     

加速器驱动次临界系统散裂靶参数优化计算

采用国际开源程序包Geant4,构建高能质子束轰击加速器驱动次临界系统(ADS)散裂靶的物理模型,模拟计算质子轰击液态金属铅、铅-铋合金和汞靶的泄漏中子谱分布,以及计算不同能量质子对应的铅靶泄漏中子产额和轴向积分分布,获得1 Ge V质子对应的铅圆柱靶优化参数,考虑入射质子的利用率和整个堆芯的体积质量,优化靶半径范围为16～24 cm,靶高为100 cm,相关研究结果可为(ADS)散裂靶的物理和工程设计提供理论依据。     

铍组件堆内长期服役后的尺寸测量

铍是核反应堆内的重要反射层材料,其辐照后的尺寸变化对反应堆的安全性具有重要的影响。为获得铍组件堆内长期服役后的尺寸变化,以对其堆内的服役性能评价提供基础数据,设计并加工了一套高放样品远程转运平台,使用三坐标测量机完成了绵阳SPRR-300堆内铍组件的尺寸变化测量实验。实验测量结果表明,SPRR-300堆的铍组件在服役29 a后,在最高中子通量高达6.78×1021 cm?2的辐照环境下,铍组件外形尺寸总体上保持良好,截面有微量的收缩变形,最大形变约0.13 mm,这表明在长期中子辐照环境下,辐照蠕变是导致铍组件尺寸变化的主要原因。     

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.     

Investigation of stability of ITER candidate Hall sensors under neutron irradiation

Hall sensors offer an attractive true non-inductive method of magnetic field measurements for fusion reactors. Their use for steady state magnetic diagnostics of ITER is presently limited by their questionable radiation and thermal stability. Issues of stable and reliable operation in ITER like radiation and thermal environment are addressed by the contribution. Recently, novel Hall sensors, compatible with temperatures up to 200°C, were developed and their radiation stability was tested at LVR-15 experimental fission reactor. Overview of the experimental set-up on LVR-15 reactor is given. Degradation of the sensor’s sensitivity by several tens of percents was observed after neutron irradiation by the total neutron fluence of 2 × 10¹⁷ n/cm² in LVR-15. This level of neutron fluence is comparable to that expected to occur over the whole ITER life time for a sensor location just outside the ITER vessel. The in-situ recalibration techniques are expected to handle the observed degree of Hall sensors performance degradation in ITER environment.     

用于BNCT-SPECT的帕维亚TRIGA MARK II反应堆热柱优化设计

针对坐落于意大利帕维亚大学的TRIGA Mark II反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源项时热柱内照射位置处中子能谱,其结果基本一致,从而验证了SSW/SSR方法的可靠性。为在该反应堆开展BNCT中SPECT实验,热柱中子束需准直为笔形束。对比分析四种热柱优化方案下束流口处及探测器处热中子和光子通量：40 cm长石墨(射束口5 cm3 cm);0.5 cm厚硼包裹40 cm长石墨(射束口5 cm3 cm);30 cm长天然锂聚乙烯(射束口直径4 cm);30 cm长天然锂聚乙烯(20 cm长射束口直径5 cm,5 cm长射束口直径4 cm,5 cm长射束口直径2 cm)。结果显示,射束口处热中子通量分别为1.05108,2.52107,6.08107和5.10 107 #/(cm2s)。综合考虑中子准直效果及光子污染,方案三具有最优性能。为后续进行BNCT-SPECT理论和实验研究提供了基础,从而有效促进BNCT剂量准确评估方法的研究进程。