Sustainability-oriented prioritization of nuclear fuel cycle transitions in China:a holistic MCDM framework under uncertainties

A sustainability-oriented assessment of the nuclear energy system can provide informative and convincing decision-making support for nuclear development strategies in China.In our previous study,four authentic nuclear fuel cycle(NFC)transi-tion scenarios were proposed,featuring different development stages and exhibiting distinct environmental,economic,and technical characteristics.However,because of the multiple and often conflicting criteria embedded therein,determining the top-priority NFC alternative for a sustainability orientation remains challenging.To address this issue,this study proposed a novel hybrid multi-criteria decision-making framework comprising fuzzy AHP,PROMETHEE GAIA,and MOORA.Initially,an improved fuzzy AHP weighting model was developed to determine criteria weights under uncertainty and investigate the influence of various weight aggregation and defuzzification approaches.Subsequently,PROMETHEE GAIA was used to address conflicts among the criteria and prioritize alternatives on a visualized k-dimensional GAIA plane.As a result,the alternative for direct recycling PWR spent fuel in fast reactors is considered the most sustainable.Furthermore,a sensitivity analysis was conducted to examine the influence of criteria weight variation and validate the screening results.Finally,using MOORA,some significant optimization ideas and valuable insights were provided to support decision-makers in shaping nuclear development strategies.

Applying multi-scale simulations to materials research of nuclear fuels:A review

Computational simulation is an important technical means in research of nuclear fuel materials.Since nuclear fuel issues are inherently multi-scopic,it is imperative to study them with multi-scale simulation scheme.At present,the development of multi-scale simulation for nuclear fuel materials calls for a more systematic approach,in which lies the main purpose of this article.The most important thing in multi-scale simulation is to accurately formulate the goals to be achieved and the types of methods to be used.In this regard,we first summarize the basic principles and applicability of the simulation methods which are commonly used in nuclear fuel research and are based on different scales ranging from micro to macro,i.e.First-Principles(FP),Molecular Dynamics(MD),Kinetic Monte Carlo(KMC),Phase Field(PF),Rate Theory(RT),and Finite Element Method(FEM).And then we discuss the major material issues in this field,also ranging from micro-scale to macro-scale and covering both pellets and claddings,with emphasis on what simulation method would be most suitable for solving each of the issues.Finally,we give our prospective analysis and understanding about the feasible ways of multi-scale integration and relevant handicaps and challenges.

Corrosion assessment for spent nuclear fuel disposal in crystalline rock,using variant cases of hydrogeological modeling

This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming computer simulations.This simplified case is presented as a base case,with changes in the hydrogeological parameters presented as variant cases.The results show that in Taiwan’s base case,decreasing the hydraulic conductivity of the rock or decreasing the hydraulic conductivity of dikes results in a shorter transport path for sulfide and an increase in corrosion depth.However,the estimated canister failure time is still over one million years in the variant cases.

Seismic considerations for spent nuclear fuel storage in dry casks

To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters. The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g. A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping. In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask. The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over). The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask. Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

Evaluation of Nuclear Fuel Centerline Temperature Using New UO2 Thermal Conductivity Models

The nuclear industry needs of prediction of behavior and life-time, for a wide range of normal, off-normal and accident conditions for safe and economic operation. Among different thermo-mechanical properties that can be predictable, the knowledge on the radial temperature distribution of the UO2 （uranium dioxide） nuclear fuel during the operation of nuclear reactors is essential for safety as different mechanical and thermal-hydraulic thresholds should be respected. One of the attributes of the Brazilian CNEN （Nuclear Energy Commission） is to assess the performance of the fuel rods used in these reactors in high-bumup regimes. The effective removal of the heat generated in the fuel rods constitutes one of the primary points to consider in the design of nuclear reactors. One of the important physical parameters in the study of heat conduction from the nuclear fuel to the coolant in a PWR （pressurized water reactor） is its thermal conductivity. It is therefore desirable that the empirical models, updated for the calculation of thermal conductivity in the fuel region be developed from new sets of experimental data from the irradiated fuel rods in controlled environments This paper presents the obtained results of implementing of a new model for thermal conductivity of the UO2 in the FRAPCON code.

Current state and prospect on the development of advanced nuclear fuel system materials:A review

The intricate balance between reactor economics and safety necessitates the emergence of new and advanced nuclear systems and,very importantly,advanced materials,which can overcome current shortcomings and bring about more economic nuclear systems with designed-in inherent safety features.These advances will achieve greater safety and better nuclear reactor economics by reaching longer reactor lives with higher levels neutron irradiation,and by providing higher operation temperatures and resistance to more aggressive corrosive environments.This paper provides a review of the current state of research and development on innovative nuclear fuel materials design and development which have the potential of benefiting simultaneously reactor economics and safety.Our discussion focuses on three areas of research:Accident-tolerant Fuels(ATFs),Oxidation Dispersion Strengthened(ODS)steels and High Entropy Alloys(HEAs).The paper also gives a prospective description of future research activities on these materials.

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 socialconcerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only underthe condition that those conflicting issues are gotten a consensus between stakeholders with different knowledgebackground, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has beendeveloped to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle systembased on the functional modeling method named Multilevel Flow Models (MFM) according to the cognition theoryof human being. Its character is that MFM models define a set of mass, energy and information flow structures onmultiple levels of abstraction to describe the functional structure of a process system and its graphical symbol representationand the means-end and part-whole hierarchical flow structure to make the represented process easy to beunderstood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system wereselected to be simulated and some analysis processes such as economics analysis, environmental analysis and energybalance 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 finallythe simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycleand its levelised cost analysis will be represented as feasible examples.

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (～22.5%), increase the energy output (～41%), decrease the quantity of spent fuels to be disposed (～2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

A method for 3D simulation of internal gas effects on thermal-mechanical behaviors in nuclear fuel elements

A new method for three-dimensional simulation of the interaction between the gas and the solid around is developed.The effects of the gas on the thermal-mechanical behaviors within the surrounded solid are performed by replacing the internal gas with an equivalent solid in the modeling,which can make it convenient to simulate the thermal-mechanical coupling effects in the solid research objects with gases in them.The applied thermal expansion coefficient,Young's modulus and Poisson's ratio of the equivalent solid material are derived.A series of tests have been conducted;and the proposed equivalent solid method to simulate the gas effects is validated.

Challenges in spent nuclear fuel final disposal:conceptual design models

<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.

Korean potential approach to the multi-lateralization of the nuclear fuel cycle

To prevent the worldwide dissemination of nuclear sensitive technologies and strengthen the safeguards of the nuclear facilities at the same time, the international society has begun to discuss the “multilateral nuclear fuel cycle approach (MNA)”. This kind of discussion will be more vigorous due to the recent nuclear activeties in Iran and North Korean and the Fukushima nuclear power plants accidents. If the MNA would be implemented someday, not even in the immediate future, Korea could be subject to a serious situation since it imports 100% of raw material for nuclear fuel. Hence, this paper reviews the 12 previous MNA proposals and discusses a potential Korean approach to MNA that Korea is able to take.

Relation between Gamma Decomposition and Powder Formation of <i>γ</i>-U8Mo Nuclear Fuel Alloys via Hydrogen Embrittlement and Thermal Shock

Gamma uranium-molybdenum alloys have been considered as the fuel phase in plate type fuel elements for material and test reactors (MTR), due to their acceptable performance under irradiation. Regarding their usage as a dispersion phase in aluminum matrix, it is necessary to convert the as cast structure into powder, and one of the techniques considered for this purpose is the hydration-dehydration (HDH). This paper shows that, under specific conditions of heating and cooling, γ-UMo fragmentation occurs in a non-reactive predominant mechanism, as shown by the curves of hydrogen absorption/desorption as a function of time and temperature. Our focus was on the experimental results presented by the addition of 8% weight molybdenum. Following the production by induction melting, samples of the alloys were thermally treated under a constant flow of hydrogen for temperatures varying from 500°C to 600°C and for times of 0.5 to 4 h. It was observed that, even without a massive hydration-dehydration process, the alloys fragmented under specific conditions of thermal treatment during the thermal shock phase of the experiments. Also, it was observed that there was a relation between absorption and the rate of gamma decomposition or the gamma phase stability of the alloy.

Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach

Nuclear fuel cell calculation is one of the most complicated steps of neutron transport problems in the reactor core. A few numerical methods use neutron flat flux (FF) approximation to solve this problem. In this approach, neutron flux spectrum is assumed constant in each region. The solution of neutron transport equation using collision probability (CP) method based on non flat flux (NFF) approximation by introducing linear spatial distribution function implemented to a simple cylindrical annular cell has been carried out. In this concept, neutron flux spectrum in each region is different each other because of an existing of the spatial function. Numerical calculation of the neutron flux in each region of the cell using NFF approach shows a fairly good agreement compared to those calculated using existing SRAC code and FF approach. Moreover, calculation of the neutron flux in each region of the nuclear fuel cell using NFF approach needs only 6 meshes which give equivalent result when it is calculated using 24 meshes in FF approach. This result indicates that NFF approach is more efficient to be used to calculate the neutron flux in the regions of the cell than FF approach.

Safe Controlled Storage of SVBR-100 Spent Nuclear Fuel in the Extended-Range Future

Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refueling equipment is used. However, comparing with RFs of nuclear submarines (NS), in which at the moment of performance of refueling the residual heat release is small, at RF SVBR-100 in a month after the reactor has been shut down, at the moment of performance of refueling the residual heat release is about 500 kW. Therefore, it is required to place the spent removable unit (SRU) with spent fuel subassemblies (SFSA) into the temporal storage tank (TST) filled with liquid LBC, in which the conditions for coolant natural circulation (NC) and heat removal via the tank vessel to the water cooling system are provided. After the residual heat release has been lowered to the level allowing transportation of the TST with SRU in the transporting-package container (TPC), it is proposed to consider a variant of TPCs transportation to the special site. On that site after the SRU has been reloaded into the long storage tank (LST) filled with quickly solidifying liquid lead, the TPCs can be stored during the necessary period. Thus, the controlled storage of LSTs is realized during several decades untill the time when SNF reprocessing and NFC closing are becoming economically expedient. On that storage, the four safety barriers are formed on the way of the release of radioactive products into the environment, namely: fuel matrix, fuel element cladding, solid lead and steel casing of the LST.

Thermal Conductivity Measurement of Zr-ZrO₂Simulated Inert Matrix Nuclear Fuel Pellet

For an evaluation of a thermal conductivity of Zr + 30 vol% ZrO2 simulated inert matrix nuclear fuel pellet, a simulated fuel pellet was fabricated using a hot-pressing method at 800°C in a vacuum and at a 20 MPa load. And several thermophysical properties of the simulated inert matrix fuel pellet were measured and calculated. The thermal diffusivity and linear thermal expansion as a function of temperature of the simulated fuel pellet were measured using a laser flash method and a dilatometry, respectively. Finally, based on the experimental data, the thermal conductivity of the simulated inert matrix fuel pellet was calculated and evaluated.

A Visual Integrated Analysis Platform for Nuclear Fuel Cycle System Based on Multilevel Flow Model

Complex Nuclear Fuel Cycle (NFC) system faces many socio-technical issues that need to obtain the consensus between stakeholders of different knowledge background. In this paper, a visualized analysis platform based on graphical functional modeling method, Multilevel Flow Model (MFM), is proposed to help those stakeholders to recognize and analyze various socio-technical issues in NFC system. Some new functions, such as “Reaction Function", “Switch Function" and “Conversion Function", are introduced to fulfill new simulation tasks for NFC system. Based upon this methodology, a micro-process and a macro-process of NFC system are simulated and meanwhile some key analysis variables, such as CO2 emission and cost flow, required by some analysis methods are deducted and displayed in the platform. And finally a sample simulation analysis is conducted based on MFM.

Study of A Multi-criteria Evaluation Methodology for Nuclear Fuel Cycle System Based on Sustainability

This paper presents a multi-criteria evaluation methodology for nuclear fuel cycle options in terms of energy sustainability. Starting from the general sustainability concept and the public acceptance questionnaire, a set of indicators reflecting specific criteria for the evaluation of nuclear fuel cycle options are defined. Particular attention is devoted to the resource utility efficiency, environmental effect, human health hazard and economic effect, which represent the different concerns of different stakeholders. This methodology also integrated a special mathematic processing approach, namely the Extentics Evaluation Method, which quantifies the human being subjective perception to provide the intuitionistic judgement and comparison for different options. The once-through option and reprocessing option of nuclear fuel cycle are examined by using the proposed methodology. The assessment process and result can give us some guidance in nuclear fuel cycle evaluation under the constraint of limited data.

Prediction of the Average Decay Heat per Fission for MOX Nuclear Fuel

MIXED Oxide Nuclear fuel (MOX) contains both uranium and plutonium in oxidized form. It is important to calculate the nuclear decay heat due to the single thermal fission (fission due to 0.0235 eV neutron) for all fissile nuclei in the MOX fuels (U235, Pu239, and Pu241). These fissile nuclei are the main source of the decay heat in MOX fuel. Decay heat calculation of the weighted fissile material content in MOX fuel is also important. A numerical method was used in this work to calculate the concentrations of all fission products due to the individual thermal fission of the three fissile materials as a function of time N(t). The decay heat calculations for the three fissile materials are directly calculated using the summation method by knowing the different concentrations of fission products over time. The average decay heat of the MOX fuel in induced thermal fission is also concluded. The most influential nuclei in the decay heat were also identified. The method used has been validated by several comparisons before, but the new in this work is using the most recent Evaluated Nuclear Data Library ENDF/B-VIII.0. Calculations of decay heat show very common trends for a period of 107 sec after the fission burst of thermal fissions of individual fissile nuclei. Moreover, the code showed high capability in calculating the fission fragments inventories and decay heats due to the decay of fission fragments of 31 fissionable nuclei.