The Spallator and APEX Nuclear Fuel Cycle a New Option for Nuclear Power

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high energy (1 to 2 GeV) protons on a heavy metal target. The neutrons are absorbed in a surrounding natural uranium or thorium blanket in which fissile Pu-239 or U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high beam current continuous wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of the short-lived fission products external to the fuel cycle eliminates the need for long-term geological age storage of fission product waste.     

Decay heat

Many aspects of the nuclear fuel cycle require accurate and detailed knowledge of the energy release rate from the decay of radioactive nuclides produced in an operating reactor. In addition to the safety assessment of nuclear power plant, decay heat estimates are needed for the evaluation of shielding requirements on fuel discharge and transport routes and for the safe management of radioactive waste products extracted from spent fuel during reprocessing. The decay heat estimates may be derived by either summation calculations or Standard equations.This paper reviews the development of these evaluation methods and traces their evolution since the first studies of the 1940s. In contrast to many of the previous reviews of this subject, both actinide and fission product evaluation methods are reviewed in parallel. Data requirements for summation calculations are examined and a summary given of available codes and their data libraries. The capabilities of present-day summation methods are illustrated through comparisons with available experimental results. Uncertainties in summation results are examined in terms of those in the basic nuclear data, irradiation details and method of calculation. The evolution of decay heat Standards is described and a brief examination made of their reliability and ability to provide suitably conservative decay heat estimates. Finally, to illustrate the use of present summation methods, comparisons are given of both the actinide and fission product decay heat levels from typical fuel samples in a variety of reactor systems.     

聚变-裂变混合堆设计研究

利用MCNP5和MONK9A程序对聚变驱动裂变混合堆进行了初步研究,在等离子体第1壁外侧依次包覆长方体形状的燃料组件和产氚组件,形成裂变堆芯包层和产氚区.对分别装载贫铀、天然铀、贫铀MOX和天然铀MOX等4种燃料的混合堆进行了研究分析,其中,后两种燃料在整个运行寿期内的功率放大倍数和氚增殖比满足设计要求.通过随燃耗变化的同位素含量分析,初步探讨了混合堆的铀-钚燃耗循环策略.     

The fuel element of the HTR-module, a prerequisite of an inherently safe reactor

To retain fission products after postulated accidents, power reactors usually rely on active safety systems inside the primary circuit, such as e.g. redundant shut down systems and multiple redundant decay heat removal systems. The HTR-Module is employing a different approach which relies entirely on the ability of the coated particle to retain all key radio-nuclides as long as a certain maximum fuel element temperature is not exceeded. Consequently, the reactor is designed such that for any postulated accident this maximum fuel element temperature is not reached even without relying on any active safety systems inside the primary circuit, since the decay heat can be removed to an outside heat sink solely by passive means.The paper discusses the experimental results of fission product release from spherical fuel elements for various temperatures. From the tests as well as from statistical considerations it can be concluded that any hazardous radiation dose to the environment can be excluded if the maximum fuel element temperature in the HTR-Module stays below 1600°C.     

球床模块式高温气冷堆失冷事故特性研究

利用高温气冷堆专用系统分析软件THERMIX程序,对球床模块式高温气冷堆(HTR-PM)失冷失压和失冷不失压事故的动态特性进行了研究,分析了堆芯功率、燃料最高温度及堆舱水冷壁余热载出功率等关键参数的变化过程,并对影响余热排出功率和燃料最高温度的不确定性进行了评价.研究结果表明,在失冷事故下,堆芯余热可通过热传导、辐射和自然对流等非能动方式传至最终热阱大气,燃料元件和压力容器等重要部件的最高温度均在设计限值内.这为HTR-PM保持模块式高温气冷堆固有安全性不变的同时实现单堆250 MW的功率方案奠定了基础,也为后续高温气冷堆电站示范工程进一步的深入设计研究提供了依据.     

聚变－裂变混合堆安全性初探

对聚变－裂变混合堆的安全性进行了初步分析和探讨．主要利用改进后的混合堆放射性程序ＦＤＫＲ对混合堆产生的核废物及放射性进行计算，并将结果与压水堆、高温气冷堆和液态金属冷却快中子增殖堆进行了比较。结果表明，混合堆与裂变动力堆相比有较好的安全性。     

Actinide reformer concept

The partitioning and transmutation technology is effective to reduce the environmental impact from disposition of high-level radioactive wastes and improve the efficiency of geological disposal. However, Am and Cm and their daughter nuclides are difficult to handle in the fuel cycle because of their high decay heat and radioactivity. These nuclides also give the chemical instability which harms the soundness of fuel pellet properties.

We propose a new system concept “actinide reformer”, which reforms Am and Cm into Pu by neutron capture reactions and decay. Am and Cm are separated from the PUREX reprocessing process and converted to chloride molten-salt fuel. Using liquid-type fuel, above mentioned defects can be compensated. Actinide reformer is an accelerator-driven system which is composed of a 10 MW-class cyclotron, a tungsten target and a subcritical core. Spent molten-salt fuel is reprocessed as an on-line fuel exchange manner to extract fission products and recover Pu to send back to a power generation cycle. The decay heat and neutron emission from the fuel with recovered Pu are smaller than those of MOX fuel with 5% of minor actinide addition. It expects no significant engineering difficulties and cost increase for construction of MOX fuel based reprocessing/fabrication plant and power reactors.     

次临界能源堆物理性能初步分析

次临界能源堆（SER）是由托卡马克聚变源驱动的聚变裂变混合堆。SER以天然铀为燃料、水为冷却剂,主要目标是生产电能。本工作建立了次临界能源堆环形圆柱模型,利用蒙特卡罗输运和燃耗计算程序,比较了燃料区不同构型对k_eff、M、TBR和燃料增殖比等参数的影响,针对均匀模型进行中子源效率与聚变源强、功率分布与能谱、初步燃耗、寿期末停堆衰变热和卸载燃料放射性等物理性能分析。计算结果表明,该模型能满足能量倍增大于6、氚自持、较长时间不换料等设计目标。研究结果为下一步开展SER安全分析提供了基础。     

Dynamic assessments on high-level waste and low- and intermediate-level waste generation from open and closed nuclear fuel cycles in Republic of Korea

We estimated the generation of low- and intermediate-level waste (LILW) and high-level waste (HLW) from open and closed nuclear fuel cycles. The closed fuel cycle reflects the development and deployment of fast reactors and pyroprocessing from 2013 to 2100, while the open fuel cycle only considers pressurized water reactors. The closed fuel cycle hardly affects short-term spent fuel management but can save nearly 60% space of interim storage compared with the open fuel cycle. Compared with the open fuel cycle, the accumulated volume of HLW can be significantly reduced in the closed fuel cycle up to over 95% in 2100. For this volume reduction, high heat generating fission products should be separated from transuranic waste in pyroprocessing and stored in decay storages for a few hundred years. Mining and milling waste in the closed fuel cycle decreases by about 31%. In contrast, the closed fuel cycle generates 3.0%–4.5% more LILW than the open fuel cycle because fast reactors and pyroprocessing produce more LILW and conversion, enrichment, and fabrication produce less LILW. In the closed fuel cycle, operation and decommissioning wastes from reactor and pyroprocessing, respectively, contribute to 74% and 8% of LILW excluding mining and milling waste.     

Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.     

高功率脉冲加热条件下燃料模块流动传热特性研究

在Z箍缩驱动的聚变-裂变混合堆中,聚变中子源以脉冲形式释放,裂变燃料在释放能量过程中具有强烈的脉冲特性。本文通过对水冷包层中的燃料模块进行抽取与建模,采用Fluent程序与集总参数法分别模拟分析了燃料模块在脉冲加热条件下的流动传热特性。采用Fluent程序对集总参数法模拟脉冲加热的可行性进行了论证,并分析了不同材料热物性对流动传热特性的影响。计算结果表明:集总参数法具有模拟脉冲加热的能力;脉冲加热条件下,流体出口温度与流量将发生振荡,且提高燃料热容与降低压力管热导率可有效降低该振荡。     

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

An application of metal fuel cycle technology toward Self-Consistent Nuclear Energy System (SCNES) concept

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and LLFPs burning capability. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

LWR spent fuel transmutation with fusion-fission hybrid reactors

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.     

Gas-cooled fusion-fission hybrid reactor systems

In this paper, the concept of the fusion-fission hybrid reactor is reviewed, and a system of classification for hybrid blanket designs is suggested. The advantages and disadvantages of gas cooling for hybrid reactor systems are discussed and the design implications of using gas cooling in a hybrid blanket are presented. Five of the more complete gas-cooled hybrid reactor conceptual design studies are discussed, and the fission-suppressed hybrid blanket concept is identified as offering potentially significant advantages in terms of inherent safety features and reduced technology development requirements compared to higher power fission blankets. It is concluded that helium is attractive as the coolant for hybrid reactor systems, and that technically viable reactor designs have been developed using helium cooling. The helium-cooled fission-suppressed hybrid blanket, based on thorium fuel for production of²³³U, is identified as being a particularly attractive candidate for further hybrid reactor development work.     

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN.

It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the ²⁴¹Pu content in the initial fuel, and the decay heat mainly depends on ²³⁸Pu and ²⁴⁴Cm. The contribution of ²⁴⁴Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from the waste disposal point of view, the characteristics of the heat generation FP elements, the platinum group metals, Mo and the long-lived FPs (LLFPs) were also investigated.     

MCORGS及次临界能源堆包层概念研究进展

次临界能源堆是以能源供应为目的的一种聚变裂变混合堆,以聚变驱动,天然铀为裂变燃料,轻水为冷却剂。本文针对该混合堆开发了基于MCNP与ORIGENS的三维中子输运燃耗耦合程序MCORGS,分析了包层三维中子学模型。提出简化干法后处理,设想利用衰变热将乏燃料加热到2 100K,将沸点低于该温度的裂变产物挥发去除。计算了包层各区材料每年发生的原子移位数,建议采用10a左右的换料周期,乏燃料经后处理后可多次复用。第1个寿期内氚增殖比TBR平均约1.15,包层能量放大倍数M平均约12;第2~9个寿期内TBR平均约1.35,M平均约18。利用流体动力学程序完成了包层CAD模型建立、网格划分及稳态传热计算分析,各区材料的最高温度均低于许用温度并有较大裕量。     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors – A case study

Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.     

聚变裂变混合堆钍基增殖锕系元素嬗变包层中子学初步研究

聚变裂变混合堆比纯聚变堆在工程及技术方面要求低,且在产生核燃料、嬗变长寿命核废料以及固有安全性方面具有一定优势,因此,越来越受到人们的重视。增殖包层是混合堆系统的关键部件,已有的包层研究基本上是基于较成熟的铀-钚燃料循环技术。针对我国铀资源相对较少而钍资源较丰富的现状,本文就一种新型的钍基燃料增殖锕系元素嬗变包层进行了初步的中子学研究,利用一维离散纵标法燃耗程序BISONC以及Monte-Carlo粒子输运程序MCNP,对包层的关键核参数,诸如氚增殖比、少量锕系元素的嬗变质量、233U产量以及热功率等,进行了较详细的计算分析。计算结果表明,生成的核燃料233U的富集度可达到3.65%,从而满足压水堆燃料富集度要求。分析结果为下一步的包层优化设计提供了依据。