The role of pyro-processing in decreasing disposal cost in Korea

A comparative analysis regarding the disposal cost of HLW (High-Level Waste) from 20,000-ton PWR nuclear fuel, focusing on pyro-processing and direct disposal, was conducted in this study. A cost estimation of the major cost drivers in disposing of pyro-processed waste revealed that canisters would cost 67.32 MEUR and that the disposal holes and disposal tunnels would require about 11.2 MEUR for excavation. These estimates amount to 1/16 and 1/55 of the costs for direct disposal of PWR spent fuels, respectively. These significant disposal cost savings in pyro-processed radioactive waste result from a significant reduction in the amount of radioactive waste to be disposed of thanks to the recycling in a fast reactor.     

Remote Fabrication of DUPIC Fuel Pellets in a Hot Cell under Quality Assurance Program

The Korea Atomic Energy Research Institute (KAERI) has been developing the Direct Use of Spent Pressurized Water Reactor (PWR) Fuel in the CANada Deuterium Uranium (CANDU) Reactors (DUPIC) fuel fabrication technology since 1992, and the basic DUPIC fuel fabrication process was established in 2002. In order to demonstrate the robustness of the DUPIC fuel fabrication process through the irradiation test, it is important that a Quality Assurance (QA) program should be in place before a fabrication of the DUPIC fuel. Therefore, the Quality Assurance Manual (QM) for the DUPIC fuel was developed on the basis of the Canadian standard, CAN3-Z299.2-85. This manual describes the quality management system applicable to the activities performed for the DUPIC fuel fabrication at KAERI. In order to demonstrate the DUPIC fuel fabrication technology and produce qualified DUPIC fuel pellets, the process qualification tests were performed, which include three pre-qualification tests and three qualification tests. The characteristics of the DUPIC fuel pellets such as the sintered density, grain size, and surface roughness were measured and evaluated in accordance with the QA procedures. The optimum fabrication process of the DUPIC fuel pellet was also established based on the qualification results. Finally a production campaign was carried out to fabricate the DUPIC fuel pellets at a batch size of 1 kg following the QA program. As a result of the production campaign, qualified DUPIC fuel pellets were successfully produced and, therefore, the remote fuel fabrication technology of the DUPIC fuel pellet was demonstrated.     

Cost efficiency analysis of disposal alternatives in Korea

This research presents the results of calculating the disposal cost efficiency for the four disposal alternatives for the CANDU spent fuel that are under development in Korea currently. The KRS-1 alternative, developed first, was set as the standard, and the efficiency of the KRS-1 alternative was assumed to be 100%.The cost calculation result shows that the A-KRS-22, which was developed most recently among the CANDU spent fuel disposal alternatives, manifested −61.7%, −45.7%, −47.0%, −78.9% and −61.7% when compared to the KRS-1 alternative concerning disposal tunnel excavation, disposal hole excavation, bentonite, disposal canister and backfilling.Moreover, the cost calculation method for the dominant cost driver that uses the unit disposal module concept for the calculation of cost efficiency was used. As for the reason that the standard for efficiency measurement was taken per each bundle, it is because the amount of bundle capacity concerning the spent fuel differs by disposal canister.     

核燃料循环成本与核电的竞争力

国际市场核燃料价格变幻莫测,对我国核电的成本和发展产生影响,本文提出了控制整个核燃料循环成本的设想,以提升中国核电的竞争力,促进核燃料产业和核电产业的共同发展。     

Availability of a probabilistic cost estimation for the price effect of Cu powder and bentonite on an HLW disposal cost in Korea

This study aims to demonstrate the availability of a probabilistic cost estimation related to the price effects of Cu powder and bentonite. From a sensitivity analysis of those materials’ prices on the overall disposal costs, it was found that Cu powder was a more dominant cost driver than that of bentonite among the material costs to dispose of 52,000 tU of spent fuels by the deterministic cost estimation method even though the used volume of Cu powder will be smaller than that of bentonite, whereas those conclusions can be changed by a probabilistic cost estimation method. Namely, its conclusion depends on a decision maker's personal opinion because of the resultant uncertainties. The disposal cost includes too many uncertainties due to the long construction and operational durations of a repository. Therefore a probabilistic cost estimation can be useful to provide the information related to an uncertainty.     

聚变驱动次临界堆经济性分析

利用自主开发的聚变系统分析软件SYSCODE对聚变驱动次临界堆的经济性进行了计算和分析。首先,对典型的环径比(标准环径比)设计方案,从堆芯几何参数、物理及工程参数等两个方面分析它们与成本的关系,然后,详细分析低、标准、高环径比三套候选设计方案的经济性与中子学参数的关系。研究发现,其经济性主要取决于包层中子学参数,另外,其经济性与几何、物理及工程参数的关系在一定条件下有别于纯聚变系统。     

The credit analysis of recycling beryllium and uranium in BeO-UO2 nuclear fuel

This study quantifies the credits of beryllium and uranium which are used as the raw materials for BeO-UO₂ nuclear fuel by analyzing the influence of their credits on the nuclear fuel cycle cost was analyzed, where the credit was defined as the value of raw materials recovered from spent fuel and the raw materials that were re-cycled. The credits of beryllium and uranium at 60 MWD/kg burn-up were –0.22 Mills/kWh and –0.14 Mills/kWh, respectively. These findings were based on the assumption that the optimal mixing proportion of beryllium in the BeO-UO₂ nuclear fuel is 4.8 wt%. In sum, the present study verified that the credits of beryllium and uranium in relation to BeO-UO₂ nuclear fuel are significant cost drivers in the cost of the nuclear fuel cycle and in estimating the nuclear fuel cycle of the reprocessing option for spent nuclear fuels.     

A cost/benefit analysis of commercial fusion-fission hybrid reactor development

A simple algorithm was developed that allows rapid computation of the ratio,R, of present worth of benefits to present worth of hybrid R&D program costs as a function of potential hybrid unit electricity cost savings, discount rate, electricity demand growth rate, total hybrid R&D program cost, and time to complete a demonstration reactor. In the sensitivity study, these variables were assigned nominal values (unit electricity cost savings of 4 mills/kW-hr, discount rate of 4%/year, growth rate of 2.25%/year, total R&D program cost of $20 billion, and time to complete a demonstration reactor of 30 years), and the variable of interest was varied about its nominal value. Results show thatR increases with decreasing discount rate and increasing unit electricity savings and ranges from 4 to 94 as discount rate ranges from 5 to 3%/year and unit electricity savings range from 2 to 6 mills/kW-hr.R increases with increasing growth rate and ranges from 3 to 187 as growth rate ranges from 1 to 3.5%/year and unit electricity cost savings range from 2 to 6 mills/kW-hr.R attains a maximum value when plotted against time to complete a demonstration reactor. The location of this maximum value occurs at shorter completion times as discount rate increases, and this optimal completion time ranges from 20 years for a discount rate of 4%/year to 45 years for a discount rate of 3%/year.The views expressed in this paper are solely those of the author and do not necessarily represent the views of the U.S. Department of Energy.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

Neutronic and thermal-hydraulic analysis of alternative ceramic fuels in the next-generation of light water reactors

The Burn-Up enlargement is one of the most important issues in the nuclear reactor core fuel management. In recent years some reactor design companies have focused on the reactor cycle length enlargement in next generation of pressurized water reactors. An increased cycle length results in an increased fuel burn-up which directly leads to low electricity costs and more efficiency. One of the promising issues is to change the chemical state of fuel that is on the agenda of the Mitsubishi Company as US-APWR nuclear power plant designer. In the present study, the neutronic as well as thermal-hydraulic analysis of some commercial ceramic fuels such as UN, UC, and UN¹⁵ instead of conventional UO₂ have been studied. The sub-channel analysis approach has been selected for these investigations. In this regard, a US-APWR fuel assembly was modelled using MCNPX2.6 Monte Carlo code by considering the periodic boundary condition in X–Y directions. It was found that the use of UC and UN¹⁵ instead of UO₂ has a deep effect on the reactor cycle length such that the power plant operational time was increased by a factor of 1.5. The COBRA-EN code with modified MATPRO subroutine has been used in thermal-hydraulic tasks. Since the thermal conductivity of these selected fuels is six times greater than UO₂, the thermal-hydraulic analysis of candidate fuels was led to outstanding results. It was found that the fuel centerline temperature in UN¹⁵ and UC cases are about half of UO₂ one, which is drastically beneficial. In summary the thermal power of next generation of pressurized water reactors could be increased considerably by using the candidate ceramic fuels instead of conventional UO₂ one.     

CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究

对压水堆乏燃料后处理回收铀（RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析。使用ORIGEN2程序．对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算。证明RU燃料元件生产的放射性水平是可以接受的。使用DRAGON／DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明：采用简单的2燃耗区，2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制。通过放射性分析和堆芯物理分析可以看出，秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下，从天然铀(NU)燃料过渡到RU燃料是可行的。通过对秦山三期CANDU堆应用RU的经济性分析，可以看出PWR／CANDU联合核燃料循环的策略既可节约铀资源(23％)，提高燃料的能量输出(4l％)．又减少了废燃料的处置量(66％)．可大大降低核电成本。     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

Economic assessment of alternative Pyro-SFR nuclear fuel cycle in Korea

In this paper, an attempt was made to present the cost required to dry-process spent fuel for a light-water reactor (PWR) and to evaluate the economy of Pyro-processing. As the Pyro-processing cost of dry-processing is calculated to be 781 $/kgHM and the break-even point of the Pyro-processing cost for direct disposal is calculated to be approximately $800/kgHM, it was determined that the Pyro-SFR connected nuclear fuel cycle is somewhat more economical than direct disposal. However, since there is no commercial Pyro facility available as of yet, uncertainty regarding the Pyro facility cost is very large. And as a result of a cost uncertainty analysis, the cost differential between direct disposal and Pyro-SFR nuclear fuel cycle has turned out to be within the statistical range of error. Hence, a judgment as to the relative economic benefit between direct disposal and Pyro-processing is hard to make for the time being. However, if a more efficient continuous process technology is developed for Pyro-processing in the future, the economical viability of the process is expected to be improved.     

Study of visualized simulation and analysis of nuclear fuel cycle system based on multilevel flow model

Complex energy and environment system, especially nuclear fuel cycle system recently raised social concerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only under the condition that those conflicting issues are gotten a consensus between stakeholders with different knowledge background, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has been developed to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle system based on the functional modeling method named Multilevel Flow Models （MFM） according to the cognition theory of human being. Its character is that MFM models define a set of mass, energy and information flow structures on multiple levels of abstraction to describe the functional structure of a process system and its graphical symbol repre- sentation and the means-end and part-whole hierarchical flow structure to make the represented process easy to be understood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system were selected to be simulated and some analysis processes such as economics analysis, environmental analysis and energy balance analysis related to those flows were also integrated to help stakeholders to understand the process of decision-making with the introduction of some new functions for the improved Multilevel Flow Models Studio, and finally the simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycle and its levelised cost analysis will be represented as feasible examples.     

Isolation of 137Cs with Copper Ferrocyanide-Anion Exchange Resin

The adsorption and desorption of ¹³⁷Cs on copper ferrocyanide-anion exchange resin, prepared in the manner previously reported in this Journal in a Short Note, are presented in detail.

This resin can also be used for the concentration of the ¹³⁷Cs found in sea water. The nuclide is adsorbed effectively on the resin from water, hydrochloric acid below 4 M, and on nitric acid below 1M. After adsorpsion on the resin, ¹³⁷Cs is eluted easily with either nitric acid (>6M), ammonium water, or silver nitrate solution.

Adsorption on the resin is specific for ¹³⁷Cs, and the action is due entirely to the ferrocyanide moiety of the resin.

This method is more efficient than the co-precipitation method with copper ferrocyanide for the concentration of radiocesium from a large volume of sea water.     

REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables

The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA at Hot Zero Power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS v2.7 a REA in Almaraz NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions.     

Numerical analysis of pressure load in a PWR cavity in an ex-vessel steam explosion

Ex-vessel steam explosion may happen as a result of melting core falling into the reactor cavity after failure of the reactor vessel and interaction with the coolant in the cavity pool. It can cause the formation of shock waves and production of missiles that may endanger surrounding structures. Ex-vessel steam explosion ener- getics is affected strongly by three dimensional （3D） structure geometry and initial conditions. Ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is developed for simulating fuel-coolant interactions. The reactor cavity with a venting tunnel is modeled based on 3D cylin- drical coordinate. A study was performed with parameters of the location of molten drop release, break size, melting temperature, cavity water subcooling, triggering time and explosion position, so as to establish parame- ters＇ influence on the fuel-coolant interaction behavior, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. The most dangerous case shows the pressure loading is above the capacity of a typical reactor cavity wall.     

Effect of Radial Peaking Factor Limitation on Discharge Burnup

Effect of the radial peaking factor limitation on the discharge burnup was examined. In general, lower limitation of the radial peaking factor places restrictions on feasible loading patterns and decreases core performance and economic efficiency. In this paper, relationship between limitation of the radial peaking factor and the discharge burnup was quantitatively investigated in 2-loop and 3-loop PWRs for several cycle lengths and fuel types. Equilibrium cores were generated assuming various radial peaking factor limitations and the change in discharge burnup, which can be considered an index of fuel cycle costs, was evaluated for each case. In order to make accurate comparisons, the generated equilibrium cores were optimized using the OPAL code by the simulated annealing method. From the calculation results, it was revealed that the limitation of the radial peaking factor has considerable impact on the discharge burnup. Relationship between the prediction accuracy of the radial peaking factor and the fuel cycle cost can be also quantitatively estimated from the above results. Therefore, the results can provide a strong motivation to improve in-core fuel management methods.     

高放废物地质处置库热学分析的参数敏感性研究

应用FLAC3D软件建立高放废物地质处置库热学分析的简化计算模型,选择影响处置库温度场的包括材料热学参数、几何参数以及时间参数在内的16个关键参数,以膨润土内表面峰值温度(该物理量是高放废物地质处置库热学设计计算中作为温度准则的物理量)为参数敏感性分析的目标物理量,通过热学计算开展参数敏感性分析。在参数敏感性分析中,将参数敏感程度划分为高、中、低三等。分析表明:4个参数(膨润土导热系数、膨润土厚度、围岩导热系数、高放废物中间贮存时间)为高敏感度参数,2个参数(散热材料厚度、回填材料厚度)为中度敏感性参数,其它10个参数(高放玻璃固化废物体、外包装容器、散热材料、回填材料的导热系数与比热,以及膨润土与围岩的比热)为低敏感度参数。通过分析可以得到如下结论:在设计高放废物地质处置库时,对膨润土及围岩导热系数的测试应力求准确,对测试结果数据认真分析,确保为设计计算提供合理的输入参数;在确保膨润土满足工艺要求功能的前提下,宜尽量减小膨润土的厚度;按照本文热学分析模型初步估算,我国高放废物至少需要中间贮存20 a以上。