BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

Interpretation of actinide transmutation in thermal and fast reactors

The commonly used transmutation rate of minor actinides in nuclear reactors is decomposed into four components, overall fission rate, Pu production rate, MA production rate, and element production rate. The physical meanings of these factors are described. The transmutation rates of minor actinides in two types of highly-moderated PWRs, a MOX fueled Na cooled fast reactor, and a metal fueled Pb cooled fast reactor are interpreted using the four components. The metal fueled Pb cooled fast reactor can incinerate minor actinides most (79kg/GWth/year), and this amount is about 4 times larger than the thermal reactors. The thermal reactors have large relative overall fission rates for ²⁴¹Am and have a potential for the incineration of ²⁴¹Am.     

Actinide recycling by pyro-process with metal fuel FBR for future nuclear fuel cycle system

Pyro-metallurgical technology is one of potential devices for future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required for the fuel cycle. In order to satisfy the requirement, actinides recycling applicable to LWR and FBR cycles by pyro-process has been developed since more than ten years in CRIEPI. The main technology is electrorefining for U and Pu separation and reductive-extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology on separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance on the geologic disposal.

The main achievements are summarized as follows:

• -|The elemental technologies, such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification, have been developed successfully with lots of experiments

• -|The fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining have been demonstrated by engineering scale facility in Argonne National Laboratory by using spent fuels and in CRIEPI by uranium tests.

• -|Single element tests, using actinides, showed the Li reduction to be technically feasible, remaining the subjects of technical feasibility on multi-elements system and on effective recycle of Li by electrolysis of Li₂O.

• -|Concerning on the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has been also established through uranium tests.

• -|It is confirmed that more than 99% of TRU nuclides can be recovered from the high level liquid waste by TRU tests

• -|Through these studies, the process flow sheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established.

The subjects to be emphasized for further development are classified into three categories, that is, process development (demonstration), technology for engineering development, and supplemental technology.

The metal fuel FBR has a high potential for recycling actinides by integration with pyro-reprocessing. Alloys of U-Pu-Zr with minor actinides are investigated from points of fuel properties. The miscibility and other characteristics suggest that the maximum content up to ca. 5 wt% of minor actinides is allowable in the matrix. Nine pins of metal fuel including minor actinides are ready for irradiation at Phenix fast reactor.     

Accident analysis on supercritical water-cooled reactor during design basis and beyond design-basis condition

Supercritical water-cooled reactor (SCWR) is the only water-cooled reactor among six Generation IV reactor concepts. Safety analysis is one of the most important tasks for SCWR design. A typical thermal spectrum SCWR with passive safety system during design-basis accident (DBA) and beyond design-basis accident (BDBA) is performed. For DBA, reactor system is modeled based on a revised code ATHLET-SC. Loss of coolant accident is chosen to perform safety analysis and sensitive analysis. The results achieved demonstrate the feasibility of proposed passive cooling system to provide sufficient cooling. However, it should be noted that if one of safety systems fails to actuate during loss of coolant accident, although the likelihood is fairly low, there is potential risk of cladding failure. Consequently, the DBA will develop into the BDBA. For BDBA, a postulated severe accident is analyzed after melt pool is formed in the lower plenum. Heat transfer behavior in the melt pool as well as two-dimensional heat transfer effect in the lower head wall is discussed. Then, key parameters are chosen to perform parametric analysis. Results show that the safety margin to critical heat flux is significant. After considering two-dimensional heat conduction effect in the lower head, the safety margin could be further increased.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B₄C or uranium–zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.     

Renewed interest in lead cooled fast reactors

In the last few years a number of compact designs of lead-alloy cooled systems have been promoted. Moreover, in Russia a design effort was started earlier on the pure lead-cooled BREST reactor but this effort does not appear to be strongly funded any more. But now the lead cooled and compact STAR-LM reactor is promoted in the US and in the European Union there is some interest in a mediumsized lead-cooled fast reactor (LFR). It has brought some nuclear industries, a large utility, several research centers and universities together to ask the European Commission for a partial funding of design and safety efforts. A 600 MWe LFR design is proposed which would be useful for base load operation but as a fast system it could also be used for load following. Because of the possible plant simplifications and the use of pure lead, the economics of such a system should be good. Moreover, efficient fuel utilization, the burning of higher actinides and a closed fuel cycle make it a sustainable system. Whether, this larger system has the same inherent / passive safety characteristics as smaller LFRs needs to be examined. In this paper the passive emergency decay heat removal by reactor vessel aircooling of such a larger system is investigated. Moreover an inlet blockage in a subassembly of a low power density LMR is analyzed. Furthermore, the pros and cons of lead vs. lead/bismuth coolants are discussed.     

铀酰与多肽和蛋白质相互作用研究进展

随着核能事业的快速发展,人们越来越关注锕系核素对生态环境和生物体所造成的影响,其中锕系元素的生物毒理行为正成为核能基础研究热点之一。锕系阳离子与生物分子特别是多肽和蛋白质的相互作用研究对于理解其在生物体内转运、吸收和沉积等基本毒理学问题至关重要。铀作为核燃料循环中主要的锕系元素,其毒理学问题更具研究意义。本文综述了铀酰离子(UO2+2)在分子水平上与氨基酸、多肽和蛋白质之间相互作用机理的研究进展。分析了所形成配合物的配位模式、分子结构以及热力学数据等;评价了血浆蛋白对铀在人体内转运、吸收和沉积所起的作用;讨论了能特异性识别UO2+2的多肽和蛋白质的设计原理,对本领域今后的发展动向也进行了展望。     

Scenarios for the transmutation of actinides in CANDU reactors

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor.Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio.This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past ( [Boczar et al., 1996] , [Chan et al., 1997] , [Hyland and Dyck, 2007] and [Hyland et al., 2009] ). The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel.The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100–1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.     

Electrodeposition of Uranium and Transuranic Elements onto Solid Cathode in LiCl-KCl/Cd System for Pyrometallurgical Partitioning

A pyrometallurgical partitioning process is being developed for recovering transuranic elements (TRUs) from high-level liquid waste. In the process, actinides are separated from fission product, especially rare earth elements (REs), by means of an electrorefining technique or a reductive-extraction technique. In this study, electrorefining experiments were carried out in LiClKCl/Cd system to recover actinides from salt bath containing actinides and REs. Uranium and neptunium could be depleted from the salt bath and recovered onto a solid cathode with high collection efficiency and high selectivity. Plutonium and americium, however, were difficult to be recovered at high current efficiency because reduction of Nd³⁺ to Nd²⁺ at about—1.7V consumed cathodic current prior to the deposition of Pu or Am. The rotation of the cathode had rather negative effect against deposition of Am and Pu in case of coexistence of much amount of Nd because Nd²⁺ was removed from the cathode surface quickly and the reaction of Nd³⁺ to Nd²⁺ was promoted. At higher current density, Pu and Am could be recovered onto solid cathode but current efficiency became too low. The result indicated that electrorefining technique in the pyro-partitioning was effective for U and Np but not for Pu and Am.     

新型锕系元素硼酸盐结构研究综述

近几年来,一系列新的锕系元素(钍、铀、镎、钚、镅、锔)硼酸盐化合物由硼酸熔融反应制备得到。这些化合物具有异常复杂的晶体结构以及非常优越的物理性质。硼酸钍具有纯无机多孔阳离子框架结构而拥有显著的阴离子交换性能,并能够选择性地将放射性核废液中的放射性核素99 Tc几乎完全提取出来,且保持高度的稳定性;硼酸铀化合物具有非常复杂的拓扑结构,且大部分结晶于非中心对称空间群;硼酸镎化合物经常显示出镎的混合价态,其中包含一个三重价态共存于同个化合物中的罕见例子,从而提供了一个崭新的轻锕系元素废料储放形式;硼酸钚提供了新奇的三价锕系元素配位环境;硼酸镅与硼酸锔同时显示出与硼酸钚与硼酸镧的显著差异,从而衍生出新的镧锕分离及锕系内部分离方略。迄今为止,关于超钚元素化合物的晶体结构与化学键的研究屈指可数,该系统研究具有非常显著的意义并且具有十足的挑战性。对此类化合物的进一步认识极有可能促进新一代核废料储放形式的探究及核燃料循环工艺的发展,进而阻止锕系放射性废物在环境中的扩散。     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Feasibility of Fast Fission System Confining Long-Lived Nuclides

The feasibility of fast fission system confining long-lived nuclides without other supporting system as synergetics for fuel sustainment and waste incineration was studied from the aspects of nuclear material balance and neutron economy. The continuous utilization of fast fission system which confines all actinides in the reactor but discharges all FP will lead to huge accumulation of radioactive wastes such as ¹²⁹I, ¹³⁵Cs, ¹⁰⁷Pd, ⁹³Zr, ⁹⁹Tc, ¹²⁶Sn and ⁷⁹Se in the far future. Then we studied the feasibility of the system that these long-lived seven FP are also confined in the reactor with actinides. In this scheme, all the long-lived nuclides to be disposed of were exposed with neutrons in the reactor and removed as different nuclides after nuclear transmutation. As the wastes stored in the repository was composed of only shorter-lived nuclides, total amount of radioactive wastes in the repository was suppressed to be less than a few tons per 3 GWt reactor.     

Metallic fuels for advanced reactors

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.     

物理-热工耦合对超临界水堆系统特性的影响分析

物理-热工耦合是超临界水堆系统分析的关键问题之一。以日本超临界水冷热堆Super LWR的堆芯设计为例,借助Dragon编制中子截面数据库,建立双群中子扩散方程计算模块,联系同时建立的热工计算模块,得到超临界水堆的物理-热工耦合计算模型。通过对比稳态与瞬态工况下耦合前、后的热工工况,分析物理-热工耦合条件下的超临界水堆系统热工特性。结果表明：在稳态工况下,物理-热工耦合将导致内、外组件堆芯功率峰值沿轴向发生明显偏移,使得部分节点的包壳温度升高,但包壳最高温度降低;在瞬态工况下,物理-热工耦合将导致堆芯包壳最高温度的发生位置有所改变。发生给水加热丧失瞬态后,在某一时刻,外部组件的包壳最高温度将转而超过内部组件的包壳最高温度。可见,物理-热工耦合对包壳最高温度的大小和发生位置均可能产生明显影响。计算分析可为超临界水堆瞬态及安全分析提供相应理论参考。     

The GDT-based fusion neutron source as driver of a minor actinides burner

At present, accelerator driven systems seem to have a good chance to play an important role in a long-term sustainable utilization of fission reactor technology. Compared with reactors, such systems offer several advantages regarding the incineration of minor actinides. However, they demand intense neutron sources. The spallation neutron source is favoured for this purpose because of its high neutron emission intensity, which seems to be attainable. The Budker Institute of Nuclear Physics Novosibirsk is developing a 14 MeV fusion neutron source on the base of the gas dynamic trap (GDT), which is primarily destined for an irradiation facility for fusion material research. The potential of this neutron source as driver of a minor actinides burner was studied by means of neutron transport calculations and compared with a spallation source. To this end, a simple model of a burner was derived from an international numerical benchmark exercise that was conducted by the Nuclear Energy Agency of the OECD. The paper presents and discusses the main results of the study and draws the conclusion that both the source strength and the efficiency of the GDT-based neutron source must be substantially increased. Moreover, advices are derived, which show that by stretching the neutron production volume and by raising the electron temperature of the GDT plasma the desired improvements could be accomplished.     

Effect of decontamination factor of recycled actinide and FP on the characteristics of SCNES

Effect of the decontamination factor (DF) of nuclides to be confined in the closed fuel cycle was examined in terms of the system characteristics. Two kinds of indices, equilibrium mass and equivalent radiotoxicity, were used to determine the recovery perfectivity of the nuclides. By using the equilibrium mass, extremely high DF had to be attained. The required values of DF were 10⁸ for LLFP and 10¹² for actinides. On the other hand, using the radiotoxicity, inferior perfectivity of recovery, DF=10⁶, could be acceptable for actinides and there is no necessity of LLFP recycling for transmutation because they only have comparable radiotoxicity with the supplied uranium to the system. The required DF which provide ignorable loss of waste to the outside of the system depends on what index we use. Generalization of index to quantify the hazard or nuisance of nuclear waste and setting the criteria remains unsettled questions.     

Accident progression within the secondary containment boundary of large LMFBRs

This study deals with the sodium spillage phenomenon as it relates to accident energetics and containment integrity. Sodium spillage has been identified as an important issue for large LMFBRs because of the large inventory of sodium present and the potential for energetic accidents. Energetic core-disruptive events leading to slug impact could open leak paths in the reactor cover and vent sodium into the secondary containment. Sodium fires in the containment building could lead to pressurization and thermal stressing of the surrounding structure and jeopardize containment integrity. The potential consequences of such a scenario have prompted the development of analytical tools to quantify the spillage process.One of the primary concerns in assessing the integrity of secondary containment is the amount and velocity of sodium which may be ejected from the primary vessel. A parametric study has been performed, the purpose of which was to study the sensitivity of sodium spillage to accident energetics. Treatment of the spillage process was accomplished with the ICECO code employing a quasi-Eulerian method. A 1000 MWe reactor, with prescribed leak paths, is modelled and analyzed during the slug impact phase. Leak paths are assumed to exist as annular penetrations in the reactor cover and as a gap at the vessel-head junction. The behavior of sodium spillage is investigated under conditions of different accident energetics, various opening sizes, and multiple leak paths, with both stationary and moving reactor covers. The relative influence of short and long term spillage is also addressed.During the transient period immediately following slug impact it was found that spillage from annular penetrations in the reactor cover is only weakly sensitive to changes in slug velocity. The same conclusion applies to spillage from a fixed gap at the vessel-head junction. Significant sensitivity of spillage to accident energetics was seen only in cases of spillage from the vessel-head junction when the reactor cover was movable. The influence of slug impact on the motion of the reactor cover leads to the conclusion that sodium spillage is most sensitive to accident energetics inasmuch as it affects the size of the leak path.     

Accident analyses and passive measures reducing the consequences of a core-melt in CAPRA/CADRA reactor cores

As part of the Combustion Améliorée du Plutonium dans les Réacteurs Avancés/Consommation D'Actinides et de Déchets dans les Réacteurs Avancés (CAPRA/CADRA) program the feasibility of reactor systems with different neutron spectra and coolants is investigated to burn plutonium and also to destruct minor actinides and long lived fission products. In this paper, we deal with reactor cores with fast spectrum and metal cooling. The design of this type of CAPRA/CADRA cores shows significant differences compared e.g. to conventional fast reactor cores. The high Pu-enrichment and the high minor actinide load have an important influence on the core meltdown behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices, which could provide an additional line of defense against severe accident development, is given. Within the CAPRA/CADRA program, also accelerator driven subcritical systems are investigated for performing the task of transmutation and incineration. In these fast neutron systems with a strong external neutron source, the kinetic behavior is different to a critical core and new strategies and measures for accident prevention have to be investigated.     

次锕系核素在铅冷快堆中的嬗变性能

乏燃料中大部分次锕系（minor actinides, MA）核素半衰期较长,对环境具有长期放射性危害。分离 嬗变技术将次锕系核素从高放废液中分离出来,并通过反应堆嬗变为短寿命或稳定核素,从而消除其放射性危害。为研究次锕系核素与燃料均匀混合、制成嬗变棒和做燃料芯块镀层装载方式下在铅冷快堆中的嬗变特性,采用MCNP和SCALE程序进行模拟计算。结果表明,三种方式下²³⁷Np、²⁴¹Am、²⁴³Am和混合次锕系核素使有效增殖因数k_eff降低,而²⁴⁴Cm和²⁴⁵Cm使k_eff升高,且²⁴⁵Cm可使k_eff大幅度增加。不同质量的混合次锕系核素装载后,三种方式下堆芯k_eff都随装载量的增加而降低,降低幅度由小到大分别为嬗变棒、均匀混合和镀层。不同次锕系核素装载量以均匀混合方式在堆芯经过550 d辐照后,²³⁷Np、²⁴¹Am和²⁴³Am嬗变率均为正值,其中²⁴¹Am嬗变率最大,而²⁴⁴Cm和²⁴⁵Cm嬗变率均为负值,²⁴⁵Cm增加明显,总的次锕系核素嬗变率为14%,可为次锕系核素在铅冷快堆中嬗变性能评价提供参考。