可燃毒物提高小型压水堆堆芯寿期研究

针对长寿期堆芯的应用需求,开展了提高小型压水堆堆芯寿期研究。以棒状燃料为对象,对不同栅格尺寸和不同可燃毒物的选取进行计算,得出小型压水堆堆芯寿期相关影响因素。通过对不同尺寸的燃料栅格进行输运 燃耗计算,得到燃耗最佳栅格尺寸。以燃耗最佳栅格尺寸建立组件,并选择转换性能好的锕系核素²⁴⁰PuO₂作为可燃毒物,利用²⁴⁰Pu吸收中子转换成易裂变核素²⁴¹Pu的特性,对堆芯实现反应性控制和寿期延长。本研究通过对燃料栅格尺寸和可燃毒物的合理选择,提高了燃料利用率,达到延长堆芯寿期的目的。     

Molten salt cooled Encapsulated Nuclear Heat Source (ENHS)-like reactors

The feasibility of designing molten-salt cooled ENHS (Encapsulated Nuclear Heat Source)-like reactor cores with Pu¹⁵N-U¹⁵N nitride fuel for high temperature applications is assessed. The cores considered have uniform fuel composition and no blanket elements and solid reflectors. They are to operate for at least 20 effective full power years without refueling, without fuel shuffling and with burnup reactivity swing less than 0.52%. Three molten-fluoride-salts: NaF(57)-BeF₂(43), ⁷LiF(66)-BeF₂(34), and LiF(46.5)-NaF(11.5)-KF(42) are considered as the coolant and six materials: SS304, Hastelloy-N, HT-9, Mn-316SS, PCA, and SiC, are considered for the structures. It is found that, neutronically, ENHS-like cores can be designed for all combinations of molten-salt coolants and structural materials considered. Relative to the reference ENHS core, the molten-salt cooled cores require significantly tighter lattice, have softer neutron spectra, significantly more negative Doppler reactivity effect, much more positive coolant temperature and void reactivity effect and smaller reactivity worth of the control elements. Of the molten salts considered, LiF-NaF-KF offers the largest p/d ratio and is most suitable for natural circulation cooling.     

Innovations in the ENHS reactor design and fuel cycle

A number of approaches were explored for improving characteristics of the encapsulated nuclear heat source (ENHS) reactor and its fuel cycle, including: increasing the ENHS module power, power density and the specific power, making the core design insensitive to the actinides composition variation with number of fuel recycling and reducing the positive void coefficient of reactivity. Design innovations examined for power increase include intermediate heat exchanger (IHX) design optimization, riser diameter optimization, introducing a flow partition inside the riser, increasing the cooling time of the LWR discharged TRU, increasing the minor actinides' concentration in the loaded fuel and split-enrichment for power flattening. Another design innovation described utilizes a unique synergism between the use of MA and the design of reduced power ENHS cores.

Also described is a radically different ENHS reactor concept that has a solid core from which heat pipes transport the fission power to a coolant circulating around the reflector. Promising features of this design concept include enhanced decay heat removal capability; no positive void reactivity coefficient; no direct contact between the fuel clad and the coolant; a core that is more robust for transportation; higher coolant temperature potentially offering higher energy conversion efficiency and hydrogen production capability.     

SFR with once-through depleted uranium breed & burn blanket

This paper assesses the feasibility of Sodium-cooled Fast Reactor (SFR) cores that have TRU recycled seeds and once-through depleted uranium blankets. The design objective of these Seed-and-Blanket (S&B) cores is to maximize the power generated by the blanket. As the blanket fuel cost is significantly lower than the cost of the seed fuel and does not need reprocessing, increasing the fraction of reactor power generated by the blanket will reduce the total fuel cycle cost and the fuel reprocessing capacity required per unit of electricity generated. The S&B core is designed to have a prolate (“cigar”) shape seed (“driver”) to maximize the fraction of neutrons that radially leak into the subcritical blanket and reduce neutron loss via axial leakage. Both seed and blanket contain multiple batches; the blanket batches are gradually shuffled inward, while one third of the fuel batches in the seed are recycled. The preliminary study found that it is possible to design the seed to accommodate a wide range of TRU conversion ratios (CR) without significantly penalizing the burnup reactivity swing. The relatively small burnup reactivity swing enables to design the S&B core to operate at longer cycles and discharge its fuel at a higher burnup relative to conventional TRU transmutation cores with identical CR. The S&B cores can generate 1000 MW_th and fit within the S-PRISM reactor vessel. The fraction of core power generated by the blanket is between 40% and 50% without exceeding the radiation damage constraint of 200 Displacements per Atom (DPA); this fraction increases when the seed is designed to have a smaller CR. These features are expected to improve the economics of SFR.     

TRU self-recycling in a high temperature modular helium reactor

The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again.In this study, a single recycling of the self-generated TRUs is considered. The UO₂ fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO₂ fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO₂ fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO₂ fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO₂-only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).     

Power flattening options for the ENHS (encapsulated nuclear heat source) core

The feasibility of power flattening while maintaining a nearly constant k_eff over the core life is assessed for the Encapsulated Nuclear Heat Source (ENHS). A couple of approaches are considered — using different fuel dimensions and using different enrichment levels across the core. Three new cores with flattened power distribution are successfully designed: Design-I uses different fuel rod diameters but uniform fuel composition; Design-II uses different fuel enrichment in the radial direction but uniform fuel rod dimensions; Design-III is similar to Design-II but uses enrichment splitting also in the axial direction. Relative to the reference ENHS core, the BOL peak-to-average channel power ratio is reduced from 1.50 to 1.15, 1.22 and 1.15 and the average discharge burnup increases by 8.5%, 27.9% and 41.2% for, respectively, Design-I, -II and -III. The corresponding burnup reactivity swings over 20 years of full power operation are 0.37%, 0.52% and 0.60% relative to 0.22% of the reference design. Design-II and -III have a negative coolant expansion reactivity defect while in the reference design this defect is positive. The radial power flattening increases the reactivity worth of the peripheral absorbers of the three new designs while the central absorber reactivity worth is reduced but their sum is nearly maintained. The newly designed cores have slightly more positive coolant void reactivity worth than the reference ENHS core.     

Radioactivity confinement capability of “PARC” concept for the future reprocessing

Aiming at TRU waste arising reduction and economical competitiveness for the future reprocessing, is proposed an advanced process concept which is named PARC (Partitioning Conundrum Key) process. Enhancement of confinement capability for long-lived nuclides in a simplified Purex process is the primary subject of this R&D project. Technologies for long-lived nuclide recovery are under development, focused on ¹⁴C and ¹²⁹I in head end, ²³⁷Np and ⁹⁹Tc in extraction, and ²⁴¹Am the daughter of ²⁴¹Pu in effluents. Those nuclides focused here are mobile in the environment and highly concerned as potential hazardous among the long-lived nuclides in spent fuels. New functions in PARC process concept are designed to mitigate the environmental impacts of reprocessing wastes and also to improve economy of reprocessing in the future.     

Neutronics and burnup analysis of an accelerator-based tru-nitride fuel transmutation system with the ATRAS code

The neutronics and burnup analyses of an accelerator-based transmutation system with tungsten target and TRU-nitride fuel were performed with a newly developed code system named ATRAS (Accelerator-based Transmutation Reactor Analysis System). The ATRAS code is an integrated code system which can perform the hadronic cascade process above 20 MeV and neutron transport and core burnup process below 20 MeV with the spallation neutron source.

The specifications of the transmutation system are investigated. The core consists of the central spallation target region and the surrounding TRU-mononitride fuel region. The core is driven by protons at an energy of 1.0 GeV. This system was also proposed as a benchmark problem in the “OECD NEA/NSC Benchmark on Physics aspects of Different Transmutation Concepts”.

According to the calculation results by the ATRAS code, higher power density and transmutation rate were achieved with nitride fuel, and the neutron spectrum was slightly harder than that of the metallic fuel system. The burnup calculation for thermal power 800 MW was also performed with the ATRAS code. It is shown that about 300 kg of TRU are transmuted annually.     

小型长寿命核能系统燃料物理性能的研究

本文在简要说明世界上小型长寿命核能系统研究现状的基础上,提出了使用钍-铀燃料和铅-铋冷却剂构造小型长寿命堆芯的设想,并为此进行了一系列燃料物理性能的研究.对于长寿命核能系统的堆芯物理设计,使反应性随燃耗变动最小非常重要,同时应该尽可能地提高堆芯的燃耗以满足长寿命运行的需求.本文使用MCNP和MCBurn程序详细计算分析了使用不同的初始驱动燃料、不同栅格、燃料成分和类型、富集度条件下,燃料栅元的燃耗反应性变化等性能,并对其进行了能谱、转换比、富集度变化等方面的分析,经过对比初步确定了使用钍-铀燃料构造长寿命堆芯的物理条件,并以此为起点构造出一个堆芯,计算给出了反应性空泡系数等安全参数.     

Power density effect on feasibility of water cooled thorium breeder reactor

Breeding is made possible by the high value of neutron regeneration ratio η for ²³³U in thermal energy region. The reactor is fueled by ²³³U–Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile ²³³U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR < 0.3. The required enrichment and conversion ratio do not change significantly caused by power density change for very tight lattice cell (MFR < 0.3), however, its strongly depends on the power density change for higher MFR (MFR ≥ 0.3). Breeding condition of all investigated power densities can be achieved for burnup ≥ 30 GW d/t at MFR = 0.3 and it requires about 3.5% of required ²³³U enrichment. Number density of ²³³Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of ²³³Pa.     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

钍基柱状高温气冷堆不同启动燃料特性初步分析

钍是一种可转换材料,将其转换成²³³U能极大提高现有核燃料资源的储量。为实现对钍的合理利用,以模块式柱状高温气冷堆GT-MHR的燃料组件作为研究对象,选取低浓缩铀、武器级钚、核反应堆级钚等作为其启动燃料。利用栅格输运计算程序DRAGON对这3种启动燃料下的钍基柱状燃料组件的寿期初中子能谱、无限增殖系数、燃耗、转换比以及²³³U和²³²Th的含量等参数进行了分析。结果表明,在易裂变物质初装量约为9%时,与低浓缩铀和武器级钚相比,核反应堆级钚作为启动燃料时组件寿期初中子能谱较硬、转换比较高;其燃耗达90 GW•d/tHM;其无限增殖系数在寿期内的波动最小;燃耗为75 GW•d/tHM时组件中²³³U存余量与²³²Th消耗量之比达0.566。     

Feasibility of a critical molten salt reactor for waste transmutation

This study assesses the feasibility of designing a Molten Salt Reactor (MSR) using the salt mixture of LiF (15 mol%), NaF (58 mol%) and BeF₂ (27 mol%) to be critical when fuelled with TRU from LWR spent fuel without exceeding the actinides solubility limit and while extracting fission products at realistic rates. The first part of the study investigated the graphite-to-MS volume ratio on the neutron balance, transmutation characteristics and graphite lifetime. It is found that a core without graphite moderator is the preferred design option; it offers the best neutron balance, most compact design and alleviated graphite lifetime problem. The second part of the study investigated sensitivity of the epithermal spectrum core to the feed composition, power density, fission products residence time and actinides loss fraction. It is found that the transmutation effectiveness improves with increasing power density and that the shorter the LWR spent fuel cooling time is, the better becomes the MSR neutron balance. The optimal MSR design offers a remarkably high transmutation capability – fissioning of as high as 99.8% of the TRU fed. The transmutation capability of the MSR is also rated in terms of final waste radiotoxicity, decay heat, spontaneous fission neutrons emission, fissile and ²³⁷Np inventory.     

增殖燃烧一体化快堆插花式倒料方案研究

增殖燃烧一体化快堆利用快堆的增殖特性,通过倒料完成从增殖组件向燃烧组件的过渡,从而实现增殖和燃烧过程的一体化。全寿期内燃烧组件提供堆芯的绝大部分功率,而在燃烧组件周围的贫铀组件则将其中的²³⁸U转化为²³⁹Pu,实现增殖功能。通过定期倒料,堆芯在一次装料后可实现长期自持临界,维持几十年的稳定运行。合理的堆芯布置与倒料方案可更好地平衡燃料的燃烧和增殖过程。插花式的堆芯布置与倒料方案是将一部分增殖组件分散布置在堆芯高通量区,保证了增殖组件的快速增殖,同时可保持堆芯在整个反应堆寿期内具有稳定的功率分布。另外,插花式堆芯布置与倒料方案最终的组件卸料燃耗是相对均衡的,所有从燃烧区倒出的组件均具有相近的燃耗,一般在250~300 GW•d/t左右。这使得增殖燃烧一体化快堆可在不进行燃料后处理的条件下,实现铀资源的高效利用。     

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.     

水样中Pu、Np、Am和Cm的联合快速分析方法

环境监测、辐射防护、核取证和核应急等领域对环境和生物样品中²³⁸Pu、²³⁹Pu、²⁴⁰Pu、²⁴¹Pu、²³⁷Np、²⁴¹Am、²⁴³Cm和²⁴⁴Cm测定的需求日渐增大。本研究提出一个自上而下串联TEVA树脂、UTEVA树脂和DGA树脂的联合、快速、可靠、可批量操作的分析方法,该方法首先通过水合氧化钛（HTO）共沉淀将待测核素从样品基质中分离,其后使用串联层析柱中的TEVA树脂柱分离纯化Pu与Np,DGA层析柱分离纯化Am与Cm。对于α放射性核素,通过CeF₃微沉淀法制备薄层α测量源,使用高分辨率α谱仪分别测量^239+240Pu、²³⁸Pu、²³⁷Np、²⁴¹Am与^243+244Cm;对于β放射性核素²⁴¹Pu,使用液体闪烁计数器测量。²³⁶Pu和²³⁴Am示踪表明该流程的化学回收率大于80%,加标实验结果表明期望值与测量值相吻合,证明了该方法的高可信度及稳定性。α谱仪测量48 h,最小可探测活度²⁴¹Am为0.40 mBq,^243+244Cm为0.33 mBq,²³⁸Pu为0.72 mBq,^239+240Pu为0.44 mBq,²³⁷Np为0.72 mBq。液闪计数器测量1 800 s,²⁴¹Pu的最小可探测活度为0.17 Bq。使用12孔真空盒同时制备12个样品,可加快制样时间,批次制样时间小于3 h,极大地降低了样品的使用量、制备时间和分析成本。     

加速器驱动系统的燃料选择

对装载不同增殖材料的现实加速器驱动系统(ADS)的安全及嬗变超铀核素特性进行研究。分别 以(U,TRU)O2和(Th,TRU)O2作为堆芯燃料,先用LAHET和MCNP程序对ADS进行稳态模拟计 算,再耦合MCNP和ORIGEN2程序计算燃耗过程中的核素密度变化。结果显示,装载钍基燃料的 ADS对超铀核素的嬗变效果较好,且在燃耗过程中其反应性和质子流强波动较小;装载铀基燃料的 ADS则具有更安全的多普勒效应和缓发中子有效份额。总体来看,如果需要堆长时间安全嬗变超铀核 素,装载钍基燃料会取得更好的效果。     

Preliminary study on feasibility of large and small water cooled thorium breeder reactor in equilibrium states

Fuel breeding is one of the essential performances for a self-sustaining reactor system which can maintains the fuel sustainability while the reactor produces energy and consumes the fissile materials during operation. Thorium cycle shows some advantageous on higher breeding characteristics in thermal neutron spectrum region as shown in the Shippingport reactor and molten salt breeder reactor (MSBR) project. In the present study, the feasibility of large and small water cooled thorium breeder reactors is investigated under equilibrium conditions where the reactors are fueled by ²³³U–Th oxide and they adopts light water coolant as moderator. The key properties such as required enrichment, breeding capability, and initial fissile inventory are evaluated. The conversion ratio and fissile inventory ratio (FIR) are used for evaluating breeding performance. The results show the feasibility of breeding for small and large reactors. The breeding performance increases with increasing power output and lower power density. The small reactor may achieve the breeding condition when the fuel pellets' power density of about 22.5 W/cm³ and burnup of about 20 GWd/t.     

熔盐堆钍铀、铀钚燃料循环核数据不确定度分析

反应堆物理设计不确定度是第4代核能系统的QMU（quantification of margins and uncertainties）有效性认证所必须的参数之一,核数据不确定度是其重要来源。基于自主开发的耦合程序BUND（burnup uncertainty of nuclear data）,将SCALE程序TRITON和TSUNAMI-3D模块耦合,完成了熔盐堆钍铀燃料循环、铀钚燃料循环核数据引起的有效增殖因数k_eff不确定度分析,并与ENDF/B-Ⅶ.1协方差数据库计算结果进行了对比。结果显示：初始时刻,两种燃料循环模式下,核数据导致的k_eff不确定度分别为0.490%和0.582%。随燃耗的增加,核数据引起的k_eff不确定度增加。寿期末,两种燃料循环模式下,对k_eff不确定度影响显著增加的反应道分别为²³⁹Pu(nubar)、(n,f)、(n,γ)、¹⁰⁵Rh(n,γ)、¹³⁵Xe(n,γ)和²³⁴U(n,γ)、¹⁴³Nd(n,γ)、^131,135Xe(n,γ)等。     

乏燃料废包壳残留物质核素分析计算

为分析计算乏燃料废包壳残留物质的核素含量，以M310型核电机组及燃料组件为分析对象，建立了乏燃料废包壳残留物质核素含量分层计算模型，用SCALE程序计算分析了²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比等主要参数随燃耗及冷却时间的变化。计算结果表明，²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比随燃耗及冷却时间的变化均可用三阶多项式拟合。本文工作为废包壳残留物质非破坏性测量方法研究提供了数据支持。