Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

UO2燃料燃耗对高放废物管理的影响研究

从乏燃料的不同燃耗引起放射性和化学组成的变化出发,分析乏燃料经后处理后的衰变热、Mo及贵金属含量对玻璃固化工艺和玻璃固化体储存的影响,计算得到了不同燃耗乏燃料制得的高放玻璃的数量。计算结果认为：对于冷却8 a的乏燃料,决定玻璃固化体包容量的不是高放主组分的热功率;对于燃耗小于40 GW•d/tU的乏燃料,决定玻璃固化体包容量的是Mo元素含量;当燃耗大于45 GW•d/tU时,贵金属含量成为决定玻璃固化体包容量的主要因素,同时UO₂燃料燃耗与高放玻璃固化体数量上存在线性关系,燃耗增加会导致高放废物玻璃固化体数量增加。随着燃耗的增加,以Mo含量及贵金属含量计算得到的玻璃固化体数量比以衰变热计算得到的玻璃固化体数量多,因此,高放废物玻璃固化前将Mo及贵金属进行分离有利于减少高放废物玻璃固化体数量。对于UO₂燃料,燃耗加深对于高放废物玻璃固化体暂存时间几乎无影响。     

乏燃料处置的必要性及其处置库环境化学行为

为了确保核燃料循环的安全性,不宜处理的乏燃料也应该同玻璃固化体一样作为高放废物进行深地质处置。本文综述了一些前期工作,归纳了空气侵入和水的辐解产生氧化性产物是导致乏燃料UO₂基体氧化溶解的主要因素;核燃料浸出实验结果显示铀和锕系镧系元素每天的浸出量是相应核素总量的1/10⁷,比裂变产物的浸出速率小一个数量级。铁金属被各国选为高放废物处置容器材料的原因是其低价格、高强度和优秀的还原能力。在最不利的地下水侵入深地质处置库、近场处置容器防腐层破损的情景下,铁容器材料表面与地下水反应产生氢气,氢气通过还原反应消耗辐解产生的氧化性自由基和分子,并能还原乏燃料表面的U(Ⅳ),大幅度减缓乏燃料的腐蚀和溶解;乏燃料中裂变产物贵金属合金颗粒对氢气有催化作用;处置容器表面铁金属能还原沉积溶解的多价态核素U(Ⅵ)、Np(Ⅴ)、Tc(Ⅶ)、Se(Ⅳ)和Se(Ⅵ)。希望本文对我国确立以铁基金属为处置容器材料的包括乏燃料在内的高放废物深地质处置概念有参考作用。     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO₂ LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO₂ of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

Progress towards a Deep Repository for Spent Nuclear Fuel in Sweden

The Swedish Nuclear Fuel and Waste Management Co. has in operation a safe and well integrated system for handling of all radioactive residues within Sweden. The existing central repository for low- and medium-level waste (SFR) and the central interim-storage facility for spent nuclear fuel (CLAB) can accommodate all the radioactive waste produced inside Sweden. Comprehensive research, development and demonstration activities are well under way for an encapsulation plant and a deep repository for spent fuel. These two facilities remain to be constructed to complete the waste management system. Siting of the deep repository is in progress with the aim of finding a suitable and accepted site. Implementation of the deep geological repository is a technical, scientific, social and political challenge. Smooth implementation must take into consideration both facts and emotions. Patience, flexibility and respect for the democratic process are important keywords. Research facilities, such as the underground Äspö Hard Rock Laboratory and the Encapsulation Laboratory, are important to promote scientific understanding as well as to demonstrate the disposal concept and technology.     

美国放射性废物最终处置政策及技术路线

反应堆安全和核废物安全处置被认为是影响今后核能事业发展的两大障碍。自１９８０年以来，美国颁布了３个关于核废物处置的政策法令，对放射性废物的安全处置的要求及责任做出了明确规定。文章介绍了美国放射性废物处置的政策、技术路线及现状。对乏燃料及高放废物的处置，美国采取了比较慎重的态度，进行了各种方案的比较，虽然已有基本轮廓，但仍在探索之中。文章还介绍了高放废物处置中存在的一些有争议的重大问题和倾向性意见。     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

The economic effects of the deferred disposal of spent fuel in Korea

The present study analyzes the economic effects concerning deferred disposal of spent fuel through long-term storage. According to the cost analysis, a scenario that a 90-year deferral of an HLW (High-Level Waste) repository construction in favor of a long-term storage of spent fuel would be economically preferable to another scenario based on the year 2040 chosen as the starting point for construction on a repository. That is, the former scenario would cost about 1/2 of the latter. This finding is an estimated result from an economic perspective only, assuming the disposal of 20,000-ton PWR spent fuel and 16,000-ton CANDU spent fuel. Still, it seems necessary to elicit proper term of storage for radioactive waste in order to comply with the so-called Polluter-Pays principle that the current generation cannot pass on its radioactive waste to the next generation.     

A study on the burn-up characteristics of inert matrix fuels

Proper disposal of minor actinides (MA), long-lived fission products (LLFPs), and transuranium element (TRU) plays a key role in the sustainable development of fission nuclear power. Adoption of inert matrix fuels (IMFs) can effectively reduce the amount of ²³⁷Np and Np element in the spent fuel of present-day commercial power reactors. In order to study the burn-up characteristics of IMFs caused by the unique composition, burn-up calculations and MA accumulation of two typical IMFs, PuO₂ + ZrO₂ + MgO and PuO₂ + ThO₂, are performed in this paper. Results indicate that k_inf at beginning of life (BOL) and reactivity drop with burn-up for PuO₂ + ZrO₂ + MgO are much larger than those of PuO₂ + ThO₂ IMF. The yields of ²³⁷Np and Np element in IMFs are two orders smaller than those of UO₂ and mixed oxide (MOX) fuels. For the same PuO₂ volume fraction and a certain burn-up, the masses of ²³⁷Np, Np element, and ²⁴¹Am for PuO₂ + ZrO₂ + MgO are smaller than those of PuO₂ + ThO₂; however, the mass of total MA is larger. IMF has high destruction efficiencies of TRU and plutonium (Pu). The results and conclusion provide basic data for the ongoing IMF design and application study.     

Storage and final disposal of spent htr fuel in the federal republic of Germany

The spent fuel treatment concept for HTR in the FRG is based on direct disposal of the fuel in a salt dome repository. Due to high burnup and good in situ fuel utilization, direct disposal offers economic advantages, especially for low enriched uranium fuel. Besides, the safety requirements can be met by simple techniques due to the special features of the HTR fuel element: coated particle fuel, stabilized in a graphite matrix with absence of any metal and the low heat production per volume unit give favourable preconditions for intermediate storage and safe disposal. Techniques for the intermediate storage are available and practised with AVR and THTR fuel. For final disposal, emplacement in boreholes, 300–600 m in depth, using simple packaging, was chosen as reference, similar to the reference concept for heat generating, medium-active waste. So far, the results of both the development and the experimental test programme underline the chosen concept.     

A non destructive technique for determination of U and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U²³³O₂ and PuO₂ have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO_2, it was difficult to monitor the percentage of U²³³O₂ through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U²³³O₂ and PuO₂ and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

高放废物地质处置黏土岩处置库围岩研究现状

世界上很多国家都对处置库的可能围岩进行了详细研究。通过对比,认为花岗岩、黏土岩、岩盐比较适合作为处置库围岩,而黏土岩由于具有自封闭性、渗透率低等其他岩石类型不可比拟的优点,因而将黏土岩作为高放废物地质处置库围岩越来越受到各国的关注。文章同时介绍了瑞士、法国、比利时等国家在黏土岩中所进行的大量研究,均认为在黏土岩中处置高放废物和乏燃料是安全的。文章还对黏土岩处置库概念设计、黏土岩处置库围岩地下实验室研究,以及我国开展黏土岩处置库研究的意义等进行了综述。