Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing <Superscript>235</Superscript>U, <Superscript>233</Superscript>U,and <Superscript>232</Superscript>Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of ²³⁵U. It operates in the open-cycle mode involving ²³³U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.     

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the ²³¹Pa–²³²U–²³³U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of ²³²U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.     

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for <Superscript>233</Superscript>U fuel production: 1. Concept of a stationary Tokamak as a neutron source

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of ²³³U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.     

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing ²³³U from ²³²Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.     

Neutronic simulation of a research reactor core of (232Th, 235U)O 2 fuel using MCNPX2.6 code

The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of ²³³U converted from ²³²Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.     

小型模块化熔盐堆钍利用方案

钍基熔盐堆（Thorium Molten Salt Reactor,TMSR）核能系统先导专项的研究目标是研发第四代裂变反应堆核能系统（即钍基熔盐堆）。为充分利用液态燃料熔盐堆的在线添料与在线燃料处理的优势,同时考虑熔盐堆的快速部署,TMSR先导专项部署了小型模块化熔盐堆。考虑燃料处理技术现状及其可能的发展方向,小型模块化熔盐堆钍利用方案采用"三步走"战略。第一阶段采用在线加料与离线处理,实现钍的成规模利用;第二阶段采用在线加料和在线处理（U）与离线处理（MA）的结合,实现钍的高效利用;第三阶段采用在线加料及在线处理全部重金属,实现钍的自持增殖利用。随着"三步走"战略的逐步实施,钍铀燃料循环模式及后处理性能稳步提高,重金属利用率得到明显改善,同时有效降低了卸料毒性。考虑燃料许可容易度和建堆时间,首先为钍利用方案第一阶段布置了三种可能的启堆燃料,分别为低富集铀、低富集铀加钍和233U加钍。计算结果显示:以低富集铀启堆时,燃料循环性能与水堆相当;以233U启堆时,燃料利用率明显高于水堆,且其放射性毒性比水堆低约2个数量级。The missions of the Thorium Molten Salt Reactor (TMSR) Nuclear Energy System are to research and develop the thorium based molten salt reactors (MSR) belonging to the fourth generation of nuclear fission reactor system. A Small modular Molten Salt Reactor (SmMSR) is deployed to make full use of the advantages of online refueling and online reprocessing and to consider the rapid deployment of MSR. An innovative "three-stage" strategy of thorium utilization based on SmMSR is proposed to take the current condition of fuel reprocessing and its future evolution. The first stage can realize the thorium utilization at a large scale with online refueling and off-line processing. The second stage can obtain efficient thorium utilization with online refueling, online processing of uranium and off-line processing of minor actinides (MAs). The third stage is implemented with self-sustaining or breeding mode with online refueling and online processing of all heavy metals. Along with the development of three stages, the utilization of heavy metals will be obviously improved and the radio-toxicity will be significantly reduced. A SmMSR is designed to achieve the goals of the first stage of thorium utilization. And three kinds of nuclear fuel cycles with different startup fuel types (i.e., low enriched uranium (LEU), thorium mixed with LEU (LEU+Th) and thorium mixed with 233U (233U+Th)) are implemented. The results show that the performance for fuel cycle containing LEU is comparable to the pressurized-water reactor (PWR). Meanwhile, the nuclear utilization for that containing 233U is much higher than PWR, and the radio-toxicity for which is lower by ~2 magnitudes than that for PWR.     

钍俘获反应率离线伽马测量方法

为了测定聚变-裂变反应堆模型钍包层中的钍俘获率以及钍-铀转化率, 探索了一种新的钍活化离线γ测量法. 利用测量²³²Th俘获反应产物²³³Th衰变链中²³³Pa衰变放出的311.98 keV 特征γ射线, 来反推计算并最终确定²³²Th(n,γ)²³³Th的反应率, 测试实验中不确定度约6% (²³³Th/²³²Th量级为10^-17情况下). 详细介绍了此方法的背景和原理方法, 并进行简单的校验实验, 证明其能够较好地得到模拟装置中的俘获率. 与瞬发γ测量法以及质谱分析法进行对比, 本方法更适合用于聚变-裂变反应堆模型钍包层中的钍俘获率以及钍-铀转化率测量, 并有望进一步测量其他相关参数.     

Inverse-beta-decay reaction in the antineutrino flux from the fragments of <Superscript>232</Superscript>Th and <Superscript>233</Superscript>U fission

The energy spectra of antineutrinos produced in the beta decay of fragments originating from ²³³U and ²³²Th fission induced by neutrons are calculated. The relevant cross sections and the spectra of positrons produced in inverse beta decay are found. This study was motivated by the hypothesis (discussed over the past decade)t hat a self-sustained chain reaction proceeds at the center of the Earth (“georeactor”). According to the author of this hypothesis, the georeactor provides energy necessary for maintaining the Earth’s magnetic field. It is ²³⁵U and, probably, ²³²Th and ²³³U that serve as a nuclear fuel in this reactor. Data obtained in the present study can be guidelines in future experiments aimed at testing the hypothesis of the georeactor and at estimating the composition of its nuclear fuel within the development of geophysical and astrophysical investigations based on the observation of antineutrino fluxes in nature.     

Cathodoluminescence of thorium in the presence of O2, CO and gas mixtures of CO-O2 and CO-H2

The cathodoluminescence (2–6 keV incident electrons) observed from thorium (111) and (533) crystal faces was recorded and analyzed for surfaces produced under various conditions. The blue luminescence observed in the presence of a partial oxygen pressure ～ 133 μPa (～10^-6 Torr) was found to consist of a broad asymmetric major band that peaked around 468 nm on which weak bands or lines were superimposed at approximately 433, 489, 502, and 534 nm. The emission was almost totally extinguished in the presence of a partial CO pressure ～ 133 μPa (～10^-6 Torr). The thorium-oxygen cathodoluminescence (CL) is interpreted as arising from the formation of ThO₂ and the excitation of luminescence centers by the incident electron beam and their subsequent decay. The major luminescence at 468 nm arises from F centers in ThO₂. The weak bands at 433 and 534 nm may arise from surface F⁺ and F centers designated as F⁺_s and F_s. The former may also be due to an OH luminescence center. The two longer wavelength lines (489, 502 nm) superimposed on the broad major band at approximately 468 nm are interpreted as arising from Pr³⁺ impurities in the thorium lattice that gave rise to fluorescence emission. The line at 468 nm also may be due in part to the fluorescence of ThO. The cathodoluminescence spectra observed in the presence of CO, and (CO+O₂) and (CO+H₂) gas mixtures were consistent with an interpretation that O₂ in the gas phase was required in order to obtain ThO₂ on and below the surface to produce significant luminescence. Auger spectroscopy showed that exposure to CO left approximately as much oxygen on the surface as in the case of O₂ but did not produce appreciable cathodoluminescence.     

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.      

Breeding of 233U in the thorium–uranium fuel cycle in VVER reactors using heavy water

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the ²³³U–²³²Th oxide fuel of water-moderated reactors with variable water composition (D₂O, H₂O) that ensures breeding of the ²³³U and ²³⁵U isotopes. The method is comparatively simple to implement.     

Measurements of Fission and Radioactive Capture Reaction Rates Inside the Fuel of the Ipen/MB-01

This work presents the measures of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO₂ at 4.3% enrichment. From its irradiation, the rate of radioactive capture and fission had been measured as a function of the radius of the pellet disk using a Ortec GMX HPGe detector. Lead collimators had been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO₂ disk is used, being inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 h under a neutron flux of 5 ×10⁸ n/cm² s. In gamma spectrometry, 10 collimators with different diameters have been used; consequently, the nuclear reactions of radioactive capture that occurs in atoms of ²³⁸U and the fission that occurs on both ²³⁵U and ²³⁸U are measured in function of 10 different regions (diameter of collimator) of the UO₂ fuel pellet disk. Nuclear fission produces different fission products such as ¹⁴³Ce with a yield fission of 5.9% which decay is monitored in this work. Corrections in geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement.     

Measurements of delayed-neutron yields from thermal-neutron-induced fission of 235U, 233U, 239Pu, and 237Np

The experimental results on delayed-neutron yields from thermal-neutron-induced fission of some actinides in the IBR-2 pulsed reactor are presented. A method of periodic irradiation without displacement of the sample was used. The measurements of delayed-neutron total yields in thermal-neutron-induced fission of ²³⁹Pu, ²³³U, and ²³⁷Np and in cold-neutron-induced fission of ²³⁵U, ²³³U, and ²³⁹Pu were carried out. All values were obtained with the use of the value of β₀ for (n _th+²³⁵U) as a reference. Precise measurements of decay curves in the time interval 5–350 ms for ²³⁵U and ²³⁹Pu were performed.     

TMSR白光中子源关键核数据实验进展

钍铀燃料循环核数据的精度和可靠性直接关系着钍基熔盐堆的安全性和经济性。目前大多数核数据都是基于铀钚燃料循环进行开发,若直接用在钍基熔盐堆上将会出现核设计不确定度较高的问题。为了提高钍基熔盐堆物理设计所需核数据的适用性,中国科学院上海应用物理研究所自行设计并建造了紧凑型的15 MeV电子加速器驱动的白光中子源(Photoneutron Source,PNS),用于开展钍铀燃料循环核数据的实验测量。该装置已通过技术验收,并进行了一系列关键核素的核数据测量,检验了现用核数据的可靠性,为相关核素的核数据评价与改进提供了基础实验数据。     

CO2 laser assisted removal of UO2 and ThO2 particulates from metal surface

Pulsed laser assisted removal of uranium dioxide and thorium dioxide particulates from stainless steel surface have been studied using a TEA CO₂ laser. Decontamination efficiency is measured as a function of laser fluence and number of pulses. Threshold fluence for the removal of UO₂ particulates has been found to be lower than that required for the removal ThO₂ particulates. Usage of a ZnSe substrate, that is transparent to the laser wavelength used here, enabled us to decouple the cleaning effect arising out of absorption in the particulates from that in the substrate and has contributed towards understanding the mechanism responsible for cleaning. The experimental observations are also corroborated by simple theoretical calculations.     

Possibility of utilizing the D-3He fuel cycle with 3He production in a spherical tokamak reactor

Possible operating regimes of a spherical tokamak reactor based on the D-³He fuel cycle with ³He production are considered. The parameters of the plasma and magnetic system are calculated for several versions corresponding to the high power efficiency (with a power gain factor in plasma of Q = 20) in a reactor with an aspect ratio of A = 1.5. According to calculations, for an axial magnetic field in vacuum of B ₀ = 2 T, a plasma radius of a = 3 m, an average 〈β〉 value of 0.53, and a plasma temperature of 〈T〉 = 48 keV, the reactor power can reach P _fus = 500 MW. In order to achieve a power of P _fus = 1500 MW in a reactor with a = 2 m, 〈β〉 = 0.36, and 〈T〉 = 40 keV, the magnetic field should be increased to B ₀ = 5 T.     

Neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

The energy dependent neutron-induced fission cross section of 233Pa has for the first time been measured directly with monoenergetic neutrons. This nuclide is an important intermediary in a thorium based fuel cycle, and its fission cross section is a key parameter in the modeling of future advanced fuel and reactor concepts. A first experiment resulted in four cross section values between 1.0 and 3.0 MeV, establishing a fission threshold in excess of 1 MeV. Significant discrepancies were found with a previous indirect experimental determination and with model estimates.     

231Pa Mössbauer resonance in Pa metal

The Mössbauer resonance of the 84.2 keV transition in ²³¹Pa has been measured for an absorber of Pa metal at 4.2 K with respect to a source of ThO₂ at 4.2 and 65 K. The resulting nuclear parameters are compared with calculations based on the Nilsson model. The electric field gradient in Pa metal is |eq_z| = (2.05 ± 0.15)} 10¹⁸ V/cm².     

Effects of channel non-orthogonality in (d,p) reactions

Coupled channel calculations for (d, p) reactions were performed in which a strong coupling between the d and p channels was assumed. Reactions investigated were ¹⁶O(d, p) ¹⁷O(2s) and ⁴⁰Ca(d, p) ⁴¹Ca(2p) at E_d =10.5 MeV and the related (d, d) and (p, p) scattering processes. The nonorthogonality of the d- and p-channels were taken into account. The results of these calculations are presented and are compared with the coupled channel calculations neglecting the channel nonorthogonality and also compared with the DWBA and optical model calculations.